WorldWideScience

Sample records for breeder reactor fuel

  1. Breeder reactor fuel fabrication system development

    International Nuclear Information System (INIS)

    Significant progress has been made in the design and development of remotely operated breeder reactor fuel fabrication and support systems (e.g., analytical chemistry). These activities are focused by the Secure Automated Fabrication (SAF) Program sponsored by the Department of Energy to provide: a reliable supply of fuel pins to support US liquid metal cooled breeder reactors and at the same time demonstrate the fabrication of mixed uranium/plutonium fuel by remotely operated and automated methods

  2. Improved fuel element for fast breeder reactor

    International Nuclear Information System (INIS)

    The invention, in which the United States Department of Energy has participated as co-inventor, relates to breeder reactor fuel elements, and specifically to such elements incorporating 'getters', hereafter designated as fission product traps. The main object of the invention is the construction of a fast breeder reactor fuel pin, free from local stresses induced in the cladding by reactions with cesium. According to the invention, the fast breeder fuel element includes a cladding tube, sealed at both ends by a plug, and containing a fissile stack and a fertile stack, characterized by the interposition of a cesium trap between the fissile and fertile stacks. The trap is effective at reactor operating temperatures in retaining and separating the cesium generated in the fissile material and preventing cesium reaction with the fertile stack. Depending on the construction method adopted, the trap may consists of a low density titanium oxide or niobium oxide pellet

  3. Fast breeder reactor fuel reprocessing in France

    International Nuclear Information System (INIS)

    Simultaneous with the effort on fast breeder reactors launched several years ago in France, equivalent investigations have been conducted on the fuel cycle, and in particular on reprocessing, which is an indispensable operation for this reactor. The Rapsodie experimental reactor was associated with the La Hague reprocessing plant AT1 (1 kg/day), which has reprocessed about one ton of fuel. The fuel from the Phenix demonstration reactor is reprocessed partly at the La Hague UP2 plant and partly at the Marcoule pilot facility, undergoing transformation to reprocess all the fuel (TOR project, 5 t/y). The fuel from the Creys Malville prototype power plant will be reprocessed in a specific plant, which is in the design stage. The preliminary project, named MAR 600 (50 t/y), will mobilize a growing share of the CEA's R and D resources, as the engineering needs of the UP3 ''light water'' plant begins to decline. Nearly 20 tonnes of heavy metals irradiated in fast breeder reactors have been processed in France, 17 of which came from Phenix. The plutonium recovered during this reprocessing allowed the power plant cycle to be closed. This power plant now contains approximately 140 fuel asemblies made up with recycled plutonium, that is, more than 75% of the fuel assemblies in the Phenix core

  4. Alternate fuel cycles for fast breeder reactors

    International Nuclear Information System (INIS)

    In this contribution to the syllabus for Subgroup 5D, a full range of alternate breeder fuel cycle options is developed and explored as to energy supply capability, resource utilizations, performance characteristics and technical features that pertain to proliferation resistance. Breeding performance information is presented for designs based on Pu/U, Pu/Th, 233 U/U, etc. with oxide, carbide or metal fuel; with lesser emphasis, heterogeneous and homogeneous concepts are presented. A potential proliferation resistance advantage of a symbiotic system of a Pu/U core, Th blanket breeder producing 233 U for utilization in dispersed LWR's is identified. LWR support ratios for various reactor and fuel types and the increase in uranium consumption with higher support ratios are identified

  5. Reprocessing of fast breeder reactor fuels in France

    International Nuclear Information System (INIS)

    The reprocessing of breeder reactor fuels is a direct technical descendant of the reprocessing of thermal reactor fuels which was developped first. The process used is in both cases the PUREX process, which consists in dissolution by nitric acid followed by selective extraction using TBP. In France, the application of this technique to breeder reactor fuels greatly benefited from the scientific and industrial experience initially acquired with metallic fuels of the MAGNOX type and then with oxide fuels of the LWR type

  6. The fast breeder reactor fuel cycle

    International Nuclear Information System (INIS)

    This paper outlines the current national fast reactor program in France and U.K and describes the increasing plant operational experience being acquired in the two countries for fuel reprocessing and the European project of a series of demonstration reprocessing plants of sufficient capacity to serve the needs of several commercially sized fast reactors. The key futures of France and U.K. programs are: fuel dismantling and pin cropping, dissolution, fuel dissolvers, liquor clarification, plutonium accountancy, solvent extraction, product preparation and packaging, wastes and emissions and fuel fabrication (initial blending, milling, pellet pressing, etc...)

  7. Fuel Summary Report: Shippingport Light Water Breeder Reactor - Rev. 2

    Energy Technology Data Exchange (ETDEWEB)

    Olson, Gail Lynn; Mc Cardell, Richard Keith; Illum, Douglas Brent

    2002-09-01

    The Shippingport Light Water Breeder Reactor (LWBR) was developed by Bettis Atomic Power Laboratory to demonstrate the potential of a water-cooled, thorium oxide fuel cycle breeder reactor. The LWBR core operated from 1977-82 without major incident. The fuel and fuel components suffered minimal damage during operation, and the reactor testing was deemed successful. Extensive destructive and nondestructive postirradiation examinations confirmed that the fuel was in good condition with minimal amounts of cladding deformities and fuel pellet cracks. Fuel was placed in wet storage upon arrival at the Expended Core Facility, then dried and sent to the Idaho Nuclear Technology and Engineering Center for underground dry storage. It is likely that the fuel remains in good condition at its current underground dry storage location at the Idaho Nuclear Technology and Engineering Center. Reports show no indication of damage to the core associated with shipping, loading, or storage.

  8. Feasibility study on the thorium fueled boiling water breeder reactor

    International Nuclear Information System (INIS)

    The feasibility of (Th,U)O 2 fueled, boiling water breeder reactor based on conventional BWR technology has been studied. In order to determine the potential use of water cooled thorium reactor as a competitive breeder, this study evaluated criticality, breeding and void reactivity coefficient in response to changes made in MFR and fissile enrichments. The result of the study shows that while using light water as moderator, low moderator to fuel volume ratio (MFR=0.5), it was possible to breed fissile fuel in negative void reactivity condition. However the burnup value was lower than the value of the current LWR. On the other hand, heavy water cooled reactor shows relatively wider feasible breeding region, which lead into possibility of designing a core having better neutronic and economic performance than light water with negative void reactivity coefficient. (authors)

  9. Multiple recycling of fuel in prototype fast breeder reactor

    Indian Academy of Sciences (India)

    G Pandikumar; V Gopalakrishnan; P Mohanakrishnan

    2009-05-01

    In a thermal neutron reactor, multiple recycle of U–Pu fuel is not possible due to degradation of fissile content of Pu in just one recycle. In the FBR closed fuel cycle, possibility of multi-recycle has been recognized. In the present study, Pu-239 equivalence approach is used to demonstrate the feasibility of achieving near constant input inventory of Pu and near stable Pu isotopic composition after a few recycles of the same fuel of the prototype fast breeder reactor under construction at Kalpakkam. After about five recycles, the cycle-to-cycle variation in the above parameters is below 1%.

  10. Fabrication of metallic fuel for fast breeder reactor

    International Nuclear Information System (INIS)

    Natural uranium oxide fuelled PHWRs comprises of first stage of Indian nuclear power programme. Liquid metal fast breeder reactors fuelled by Pu (from PHWR's) form the second stage. A shorter reactor doubling time is essential in order to accelerate the nuclear power growth in India. Metallic fuels are known to provide shorter doubling times, necessitating to be used as driver fuel for fast breeder reactors. One of the fabrication routes for metallic fuels having random grain orientation, is injection casting technique. The technique finds its basis in an elementary physical concept - the possibility of supporting a liquid column within a tube, by the application of a pressure difference across the liquid interface inside and outside the tube. At AFD, BARC a facility has been set-up for injection casting of uranium rods in quartz tube moulds, demoulding of cast rods, end-shearing of rods and an automated inspection system for inspection of fuel rods with respect to mass, length, diameter and diameter variation along the length and internal and external porosities/voids. All the above facilities have been set-up in glove boxes and have successfully been used for fabrication of uranium bearing fuel rods. The facility has been designed for fabrication and inspection of Pu-bearing metallic fuels also, if required

  11. Development of fuels and structural materials for fast breeder reactors

    Indian Academy of Sciences (India)

    Baldev Raj; S L Mannan; P R Vasudeva Rao; M D Mathew

    2002-10-01

    Fast breeder reactors (FBRs) are destined to play a crucial role inthe Indian nuclear power programme in the foreseeable future. FBR technology involves a multi-disciplinary approach to solve the various challenges in the areas of fuel and materials development. Fuels for FBRs have significantly higher concentration of fissile material than in thermal reactors, with a matching increase in burn-up. The design of the fuel is an important aspect which has to be optimised for efficient, economic and safe production of power. FBR components operate under hostile and demanding environment of high neutron flux, liquid sodium coolant and elevated temperatures. Resistance to void swelling, irradiation creep, and irradiation embrittlement are therefore major considerations in the choice of materials for the core components. Structural and steam generator materials should have good resistance to creep, low cycle fatigue, creep-fatigue interaction and sodium corrosion. The development of carbide fuel and structural materials for the Fast Breeder Test Reactor at Kalpakkam was a great technological challenge. At the Indira Gandhi Centre for Atomic Research (IGCAR), advanced research facilities have been established, and extensive studies have been carried out in the areas of fuel and materials development. This has laid the foundation for the design and development of a 500 MWe Prototype Fast Breeder Reactor. Highlights of some of these studies are discussed in this paper in the context of our mission to develop and deploy FBR technology for the energy security of India in the 21st century.

  12. Experience with EBR-II [Experimental Breeder Reactor] driver fuel

    International Nuclear Information System (INIS)

    The exceptional performance of Experimental Breeder Reactor-II (EBR-II) metallic driver fuel has been demonstrated by the irradiation of a large number of elements under steady-state, transient overpower, and loss-of-flow conditions. High burnup with high reliability has been achieved by a close coupling of element design and materials selection. Quantification of reliability has allowed full utilization of element lifetime. Improved design and duct materials currently under test are expected to increase the burnup from 8 to 14 at.%

  13. Fuel Summary Report: Shippingport Light Water Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Illum, D.B.; Olson, G.L.; McCardell, R.K.

    1999-01-01

    The Shippingport Light Water Breeder Reactor (LWBR) was a small water cooled, U-233/Th-232 cycle breeder reactor developed by the Pittsburgh Naval Reactors to improve utilization of the nation's nuclear fuel resources in light water reactors. The LWBR was operated at Shippingport Atomic Power Station (APS), which was a Department of Energy (DOE) (formerly Atomic Energy Commission)-owned reactor plant. Shippingport APS was the first large-scale, central-station nuclear power plant in the United States and the first plant of such size in the world operated solely to produce electric power. The Shippingport LWBR was operated successfully from 1977 to 1982 at the APS. During the five years of operation, the LWBR generated more than 29,000 effective full power hours (EFPH) of energy. After final shutdown, the 39 core modules of the LWBR were shipped to the Expended Core Facility (ECF) at Naval Reactors Facility at the Idaho National Engineering and Environmental Laboratory (INEEL). At ECF, 12 of the 39 modules were dismantled and about 1000 of more than 17,000 rods were removed from the modules of proof-of-breeding and fuel performance testing. Some of the removed rods were kept at ECF, some were sent to Argonne National Laboratory-West (ANL-W) in Idaho and some to ANL-East in Chicago for a variety of physical, chemical and radiological examinations. All rods and rod sections remaining after the experiments were shipped back to ECF, where modules and loose rods were repackaged in liners for dry storage. In a series of shipments, the liners were transported from ECF to Idaho Nuclear Technology Engineering Center (INTEC), formerly the Idaho Chemical Processing Plant (ICPP). The 47 liners containing the fully-rodded and partially-derodded core modules, the loose rods, and the rod scraps, are now stored in underground dry wells at CPP-749.

  14. Fuel Summary Report: Shippingport Light Water Breeder Reactor

    International Nuclear Information System (INIS)

    The Shippingport Light Water Breeder Reactor (LWBR) was a small water cooled, U-233/Th-232 cycle breeder reactor developed by the Pittsburgh Naval Reactors to improve utilization of the nation's nuclear fuel resources in light water reactors. The LWBR was operated at Shippingport Atomic Power Station (APS), which was a Department of Energy (DOE) (formerly Atomic Energy Commission)-owned reactor plant. Shippingport APS was the first large-scale, central-station nuclear power plant in the United States and the first plant of such size in the world operated solely to produce electric power. The Shippingport LWBR was operated successfully from 1977 to 1982 at the APS. During the five years of operation, the LWBR generated more than 29,000 effective full power hours (EFPH) of energy. After final shutdown, the 39 core modules of the LWBR were shipped to the Expended Core Facility (ECF) at Naval Reactors Facility at the Idaho National Engineering and Environmental Laboratory (INEEL). At ECF, 12 of the 39 modules were dismantled and about 1000 of more than 17,000 rods were removed from the modules of proof-of-breeding and fuel performance testing. Some of the removed rods were kept at ECF, some were sent to Argonne National Laboratory-West (ANL-W) in Idaho and some to ANL-East in Chicago for a variety of physical, chemical and radiological examinations. All rods and rod sections remaining after the experiments were shipped back to ECF, where modules and loose rods were repackaged in liners for dry storage. In a series of shipments, the liners were transported from ECF to Idaho Nuclear Technology Engineering Center (INTEC), formerly the Idaho Chemical Processing Plant (ICPP). The 47 liners containing the fully-rodded and partially-derodded core modules, the loose rods, and the rod scraps, are now stored in underground dry wells at CPP-749

  15. Fast Breeder Reactor studies

    International Nuclear Information System (INIS)

    This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts

  16. Fast Breeder Reactor studies

    Energy Technology Data Exchange (ETDEWEB)

    Till, C.E.; Chang, Y.I.; Kittel, J.H.; Fauske, H.K.; Lineberry, M.J.; Stevenson, M.G.; Amundson, P.I.; Dance, K.D.

    1980-07-01

    This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts.

  17. Embattled breeder reactor

    International Nuclear Information System (INIS)

    A commercial fuel-cloning machine, a nuclear breeder reactor, is yet to produce electricity in the United States. It is expensive in capital and fuel costs, its fuel that must be reprocessed can become a link to nuclear weapons manufacture, and its safety is no greater than conventional nuclear reactors. The breeder has had on-again/off-again administrative support from Washington. Opponents worry about escalating costs and failure to develop alternatives like solar energy. Proponents say fossil-fuel depletion will eventually force long-term renewable resources such as the breeder anyway. Some who share parts of both views oppose present policy regarding the Clinch River Breeder demonstration plant specifically. The correct choices on breeder concept development and commercialization will be known in 2050. 3 figures

  18. Fast breeder reactor

    International Nuclear Information System (INIS)

    The fluid-cooled fast breeder reactor described includes an outer cylindrical boundary wall, a plurality of canless fuel elements and breeder material elements received within the boundary wall and being in an array therein forming a fissionable fuel zone and a breeder material zone coaxially surrounding the fissionable fuel zone, a coolant supply system for applying fluid coolant at uniform pressure to the entire cross section within the cylindrical boundary wall, and flow guide devices extending substantially horizontally and disposed at different levels one above the other within the breeder material zone which coaxially surrounds the fissionable fuel zone, means for elastically securing the flow guide devices at alternate levels within the breeder material to the boundary wall, the flow guide devices at the levels intermediate the alternate levels being spaced by an annular gap from the boundary wall. 7 claims, 7 drawing figures

  19. Remote fabrication of pellet fuels for United States breeder reactors

    International Nuclear Information System (INIS)

    Goal of the program is to demonstrate the feasibility of fabricating breeder fuel in a remotely operated and maintained mode by 1985. Development for pellet fuel fabrication is in the engineering stage with much of the equipment for ceramic unit operations in final design or currently under testing. Results to date confirm that remote fabrication of pellet fuels is feasible. Several of the processes and equipment items are described in this report

  20. Breeder reactors

    International Nuclear Information System (INIS)

    The reasons for the development of fast reactors are briefly reviewed (a propitious neutron balance oriented towards a maximum uranium burnup) and its special requirements (cooling, fissile material density and reprocessing) discussed. The three stages in the French program of fast reactor development are outlined with Rapsodie at Cadarache, Phenix at Marcoule, and Super Phenix at Creys-Malville. The more specific features of the program of research and development are emphasized: kinetics and the core, the fuel and the components

  1. Comparative assessment of nuclear fuel cycles. Light-water reactor once-through, classical fast breeder reactor, and symbiotic fast breeder reactor cycles

    International Nuclear Information System (INIS)

    The object of the Alternative Nuclear Fuel Cycle Study is to perform comparative assessments of nuclear power systems. There are two important features of this study. First, this evaluation attempts to encompass the complete, integrated fuel cycle from mining of uranium ore to disposal of waste rather than isolated components. Second, it compares several aspects of each cycle - energy use, economics, technological status, proliferation, public safety, and commercial potential - instead of concentrating on one or two assessment areas. This report presents assessment results for three fuel cycles. These are the light-water reactor once-through cycle, the fast breeder reactor on the classical plutonium cycle, and the fast breeder reactor on a symbiotic cycle using plutonium and 233U as fissile fuels. The report also contains a description of the methodology used in this assessment. Subsequent reports will present results for additional fuel cycles

  2. IAEA note on multi-national fuel cycle centres as related to fast breeder reactors

    International Nuclear Information System (INIS)

    The significant aspects of associating fast breeder reactor fuel cycles with the concept of regional fuel cycle centres, as studied earlier by the IAEA, are identified. The results of the RFCC Study Project are presented, and how in particular non-proliferation and safeguards, radioactive waste management and economic considerations would be effected by inclusion of fast breeder reactor fuel cycle facilities and possibly fast breeder reactors as well in such centres, are discussed. The current effort of the IAEA to develop a computer programme which models the material flows in the nuclear fuel cycle which could be applied to the analysis of alternative siting strategies for FBR and its fuel cycle facilities is discussed

  3. Plutonium bearing oxide fuels for recycling in thermal reactors and fast breeder reactors

    International Nuclear Information System (INIS)

    Programs carried out in the past two decades have established the technical feasibility of using plutonium as a fuel material in both water-cooled power reactors and sodium-cooled fast breeder reactors. The problem facing the technical community is basically one of demonstrating plutonium fuel recycle under strict conditions of public safety, accountability, personnel exposure, waste management, transportation and diversion or theft which are still evolving. In this paper only technical and economic aspects of high volume production and the demonstration program required are discussed. This paper discusses the role of mixed oxide fuels in light water reactors and the objectives of the LMFBR required for continual growth of nuclear power during the next century. The results of studies showing the impact of using plutonium on uranium requirements, power costs, and the market share of nuclear power are presented. The influence of doubling time and the introduction date of LMFBRs on the benefits to be derived by its commercial use are discussed. Advanced fuel development programs scoped to meet future commerical LMFBR fuel requirements are described. Programs designed to provide the basic technology required for using plutonium fuels in a manner which will satisfy all requirements for public acceptance are described. Included are the high exposure plutonium fabrication development program centered around the High Performance Fuels Laboratory being built at the Hanford Engineering Development Laboratory and the program to confirm the technology required for the production of mixed oxide fuels for light water reactors which is being coordinated by Savannah River Laboratories

  4. Production of nuclear fusion reactor fuel by ceramic tritium breeder material

    International Nuclear Information System (INIS)

    Fuel tritium is generated from the nuclear reaction between the fusion neutron and the lithium of the breeder material arranged in the blanket that encloses the fusion plasma in the fusion reactor. However, the release process of the generated tritium has not been completely clarified. Recently, Japan Atomic Energy Agency started the tritium generation and recovery experiment in using nuclear fusion neutron source (FNS). In this report, the recent results of study on breeder material and its manufacturing technology is presented. (author)

  5. Analysis of principle possibilities of intermediare storage of fast breeder reactor fuel elements

    International Nuclear Information System (INIS)

    The principle possibilities of intermediate storage of fast breeder reactor fuel elements were analyzed and compared on the basis of 4 different concepts of storage. The SNR-2 fuel element was chosen as reference. Only the pool (wet) storage could be used to store fuel elements of less than 18 months precooling time. The other concepts (dry storage and container storage) have distinct advantages at precooling times longer than 18 months. (orig./HP) With 22 tabs., 8 figs

  6. Development of metallic fuels for Indian Fast Breeder Reactors

    International Nuclear Information System (INIS)

    The neutronic performance of metal fuel based on binary U-Pu alloy or ternary U-Pu-Zr alloys are better than conventional uranium plutonium mixed oxide or high density carbide ceramic fuel. The growing energy demand in India needs faster growth of nuclear power and warrants introduction of fast reactors based on metallic fuels in future. Physics calculation showed that fast reactor based on metallic fuels results in higher breeding ratio and lower doubling time compare to mixed oxide or carbide fuels. Moreover inclusion of pyro-processing of the fuel in the fuel cycle is expected to make metal fuel option more economical. As part of metal fuel development programme for future FBR's in India, capsule irradiation of metal fuel based on sodium bonded U-Pu-Zr alloy and metal (Zircaloy) bonded binary U-Pu (Pu ∼ 15 %) alloy are being actively pursued. For this purpose two design concepts have been proposed : one based on sodium bonded ternary alloy fuel of U-Pu-Zr (2-10 wt%) in modified T91 cladding material and the other is U-Pu binary alloy mechanically bonded to modified T91 cladding material with 'Zircaloy' as a liner between the fuel alloy and the clad. The Zircaloy liner act as a barrier in reducing the fuel clad chemical interaction. It also helps in transfer of heat from the fuel to the clad. The smear density of metal bonded pin will be between 70% - 85% and that for sodium bonded pin will be ∼ 70%. In metal bonded fuel pin design two/four semi-circular grooves of diameter ∼1.0 mm, will be provided in diametrically opposite directions in the fuel cross section to accommodate fuel swelling. A comparison has been made on the relative merits and demerits of these two fuel pin designs. The material for the axial blanket will be 'U' or U-Zr (Zr upto 10wt %) alloy based on the results of the out-of-pile thermal cycling behavior and irradiation performance. In the present investigation out-of-pile experiments have been carried out to address some of the issues of

  7. End-of-life destructive examination of light water breeder reactor fuel rods (LWBR Development Program)

    International Nuclear Information System (INIS)

    Destructive examination of 12 representative Light Water Breeder Reactor fuel rods was performed following successful operation in the Shippingport Atomic Power Station for 29,047 effective full power hours, about five years. Light Water Breeder Reactor fuel rods were unique in that the thorium oxide and uranium-233 oxide fuel was contained within Zircaloy-4 cladding. Destructive examinations included analysis of released fission gas; chemical analysis of the fuel to determine depletion, iodine, and cesium levels; chemical analysis of the cladding to determine hydrogen, iodine, and cesium levels; metallographic examination of the cladding, fuel, and other rod components to determine microstructural features and cladding corrosion features; and tensile testing of the irradiated cladding to determine mechanical strength. The examinations confirmed that Light Water Breeder Reactor fuel rod performance was excellent. No evidence of fuel rod failure was observed, and the fuel operating temperature was low (below 25800F at which an increased percentage of fission gas is released). 21 refs., 80 figs., 20 tabs

  8. Studies of the restructuring of fast breeder test reactor fuel by out-of-pile simulation

    International Nuclear Information System (INIS)

    The fast breeder test reactor (FBTR) at Kalpakkam, India, currently employs a mixed carbide of uranium and plutonium with a Pu/(Pu + U) ratio of 0.70 as fuel. The behavior of this fuel in a thermal gradient is investigated. An out-of-pile simulation facility is designed, set up, and commissioned. Experiments are conducted on FBTR fuel pellets to study the restructuring of the fuel at various levels of linear power and its cracking behavior in a thermal gradient. The results are discussed in terms of their significance for reactor operation

  9. Fuel pins and core response under liquid-metal fast breeder reactor transient overpower accident conditions

    International Nuclear Information System (INIS)

    Since the earlier liquid-metal fast breeder reactor transient overpower assessments were done (1975), new experimental data and modeling improvements have occurred that indicate later failures and more molten fuel squirted into the channel with a higher propensity for plugging. An initial sweepout still occurs, and an analysis shows that even if coherent instead of the expected stochastic failures occur, the blockages are partial, the reactor is strongly shut down, and a coolable geometry exists. Hence, the overall consequences would be benign

  10. Helium Leak Detection of Vessels in Fuel Transfer Cell (FTC) of Prototype Fast Breeder Reactor (PFBR)

    Science.gov (United States)

    Dutta, N. G.

    2012-11-01

    Bharatiya Nabhikiya Vidyut Nigam (BHAVINI) is engaged in construction of 500MW Prototype Fast Breeder Reactor (PFBR) at Kalpak am, Chennai. In this very important and prestigious national programme Special Product Division (SPD) of M/s Kay Bouvet Engg.pvt. ltd. (M/s KBEPL) Satara is contributing in a major way by supplying many important sub-assemblies like- Under Water trolley (UWT), Airlocks (PAL, EAL) Container and Storage Rack (CSR) Vessels in Fuel Transfer Cell (FTC) etc for PFBR. SPD of KBEPL caters to the requirements of Government departments like - Department of Atomic Energy (DAE), BARC, Defense, and Government undertakings like NPCIL, BHAVINI, BHEL etc. and other precision Heavy Engg. Industries. SPD is equipped with large size Horizontal Boring Machines, Vertical Boring Machines, Planno milling, Vertical Turret Lathe (VTL) & Radial drilling Machine, different types of welding machines etc. PFBR is 500 MWE sodium cooled pool type reactor in which energy is produced by fissions of mixed oxides of Uranium and Plutonium pellets by fast neutrons and it also breeds uranium by conversion of thorium, put along with fuel rod in the reactor. In the long run, the breeder reactor produces more fuel then it consumes. India has taken the lead to go ahead with Fast Breeder Reactor Programme to produce electricity primarily because India has large reserve of Thorium. To use Thorium as further fuel in future, thorium has to be converted in Uranium by PFBR Technology.

  11. Helium Leak Detection of Vessels in Fuel Transfer Cell (FTC) of Prototype Fast Breeder Reactor (PFBR)

    International Nuclear Information System (INIS)

    Bharatiya Nabhikiya Vidyut Nigam (BHAVINI) is engaged in construction of 500MW Prototype Fast Breeder Reactor (PFBR) at Kalpak am, Chennai. In this very important and prestigious national programme Special Product Division (SPD) of M/s Kay Bouvet Engg.pvt. ltd. (M/s KBEPL) Satara is contributing in a major way by supplying many important sub-assemblies like- Under Water trolley (UWT), Airlocks (PAL, EAL) Container and Storage Rack (CSR) Vessels in Fuel Transfer Cell (FTC) etc for PFBR. SPD of KBEPL caters to the requirements of Government departments like – Department of Atomic Energy (DAE), BARC, Defense, and Government undertakings like NPCIL, BHAVINI, BHEL etc. and other precision Heavy Engg. Industries. SPD is equipped with large size Horizontal Boring Machines, Vertical Boring Machines, Planno milling, Vertical Turret Lathe (VTL) and Radial drilling Machine, different types of welding machines etc. PFBR is 500 MWE sodium cooled pool type reactor in which energy is produced by fissions of mixed oxides of Uranium and Plutonium pellets by fast neutrons and it also breeds uranium by conversion of thorium, put along with fuel rod in the reactor. In the long run, the breeder reactor produces more fuel then it consumes. India has taken the lead to go ahead with Fast Breeder Reactor Programme to produce electricity primarily because India has large reserve of Thorium. To use Thorium as further fuel in future, thorium has to be converted in Uranium by PFBR Technology.

  12. Fast breeder reactor fuel reprocessing R and D: technological development for a commercial plant

    International Nuclear Information System (INIS)

    The technological developments undertaken by the CEA are applied to a plant project of a 50 t/y capacity, having to reprocess in particular the SUPERPHENIX 1 reactor fuel. French experience on fast breeder reactor fuel reprocessing is presented, then the 50 t/y capacity plant project and the research and development installations. The R and D programs are described, concerning: head-end operations, solvent extractions, Pu02 conversion and storage, out-of-specification Pu02 redissolution, fission products solution vitrification, conditioning of stainless steel hulls by melting, development of remote operation equipments, study of corrosion and analytical problems

  13. Conceptual design of the Clinch River Breeder Reactor spent-fuel shipping cask

    International Nuclear Information System (INIS)

    Details of a baseline conceptual design of a spent fuel shipping cask for the Clinch River Breeder Reactor (CRBR) are presented including an assessment of shielding, structural, thermal, fabrication and cask/plant interfacing problems. A basis for continued cask development and for new technological development is established. Alternates to the baseline design are briefly presented. Estimates of development schedules, cask utilization and cost schedules, and of personnel dose commitments during CRBR in-plant handling of the cask are also presented

  14. Conceptual design of the Clinch River Breeder Reactor spent-fuel shipping cask

    Energy Technology Data Exchange (ETDEWEB)

    Pope, R B; Diggs, J M [eds.

    1982-04-01

    Details of a baseline conceptual design of a spent fuel shipping cask for the Clinch River Breeder Reactor (CRBR) are presented including an assessment of shielding, structural, thermal, fabrication and cask/plant interfacing problems. A basis for continued cask development and for new technological development is established. Alternates to the baseline design are briefly presented. Estimates of development schedules, cask utilization and cost schedules, and of personnel dose commitments during CRBR in-plant handling of the cask are also presented.

  15. Reprocessing technology of liquid metal cooled fast breeder reactor fuel

    International Nuclear Information System (INIS)

    All the important aspects of LMFBR fuel reprocessing are critically reviewed in this report. Storage and transportation techniques using sodium, inert gas, lead, molten salts and organic coolants are comparatively discussed in connection with cooling time and de-activation techniques. Decladding and fuel disaggregation of UO2-PuO2 fuel are reviewed according to the present state of R and D in the main nuclear powers. Strong emphasis is put on on voloxidation, mechanical pulverization and molten salt disaggregation in connection with volatilization of gaseous fission products. Release of fission gases and the resulting off-gas treatment are discussed in connection with cooling time, burn up and dissagregation techniques. The review is limited to tritium, iodine xenon-krypton and radioactive airborne particulates. Dissolution, solvent extraction and plutonium purification problems specifically connected to LMFBR fuel are reviewed with emphasis on the differences between LWR and fast fuel reprocessing. Finally the categories of wastes produced by reprocessing are analysed according to their origin in the plant and their alpha emitters content. The suitable waste treatment techniques are discussed in connection with the nature of the wastes and the ultimate disposal technique. (author)

  16. Development of sodium facilities for NSRR fast breeder reactor fuel tests. 2. Sodium capsule

    International Nuclear Information System (INIS)

    In order to commercialize fast reactors, which are expected to be long-term transmutes of plutonium and long half life radioactive wastes (such as americium) from light water reactors, safety research under accident conditions and establishment of the safety guidelines are essential. Sodium facilities, such as, (1) Purification/charging loop and test loop, and (2) Proto-type Sodium capsule, were developed and fabricated in order to pulse irradiate fast breeder reactor fuels in the Nuclear Safety Research Reactor (NSRR) of JAERI for investigation on fuel behavior under transient over-power conditions. This report presents the purpose, outlines, specifications, capabilities and operation results of the proto-type sodium capsule. Two kinds of capsule, i.e., the stagnant sodium capsule and the sodium loop, were designed to pulse irradiate Fast Reactor (FR) fuels in the NSRR under sodium cooling conditions with and without flow, respectively. Because the capsules have to safely contain chemically active sodium at high temperature and stand the pressure pulses by the sodium hummer which might be generated at fuel failure, the development of the capsule is essential for realizing the research. Thus, proto-type sodium loop, which consisted of doubly sealed container, sodium pump and flow meter, was developed. In addition, two type of flange structure for the stagnant capsule and loop was leak tested at high pressures, in order to confirm its sealing capability at room and high temperature conditions. (author)

  17. Fabrication and quality control of MOX fuel for Prototype Fast Breeder Reactor (PFBR)

    International Nuclear Information System (INIS)

    Uranium-Plutonium mixed oxide (MOX) fuel for both thermal and fast reactors have been fabricated by Advanced Fuel Fabrication Facility (AFFF), Bhabha Atomic Research Centre, Tarapur, India. MOX fuel bundles fabricated by AFFF have been loaded in Boiling Water Reactors (BWRs) and Pressurised Heavy Water Reactors (PHWRs) and have been discharged after successful irradiation. An experimental fuel subassemby containing 37 MOX pins is being irradiated in Fast Breeder Test Reactor (FBTR) at Kalpakkam near Chennai and has seen a burn up of more than 92000 MWd/t. MOX fuel pins containing 44% PuO2 have been recently loaded as a part of the hybrid core of FBTR. AFFF has now taken up the manufacture of MOX fuel pins for the Prototype Fast Breeder Reactor (PFBR) coming up at Kalpakkam . The core consists of 181 sub assemblies containing 217 MOX fuel pins each. Prototype Fast Breeder Reactor is designed with two different fissile enrichment zones to be loaded with MOX subassemblies with a nominal composition of 21% and 28% of PuO2.The fuel pellets of required composition are made using conventional powder metallurgy processes. The pellets are annular with an inner hole of 1.8 mm diameter and outside diameter of 5.5 mm. AFFF has developed the technology of making annular MOX fuel pellets for PFBR and optimized conditions of fabrication. Multistaion rotary presses have been used for compaction of the pellets. The fuel pin consists of a MOX stack of 1000 mm and axial blanket of deeply depleted uranium dioxide of length 300 mm on either side. New techniques have been used at different stages of fabrication of the fuel pins namely pelletisation, welding and wire wrapping. Studies have been made to use laser welding technique for welding of endplugs. Automation has been introduced in a number of process steps in the fabrication line. A detailed quality control plan is prepared based on the specifications and advanced process and quality control procedures have been incorporated to

  18. Fabrication and quality control of MOX fuel for Prototype Fast Breeder Reactor (PFBR)

    International Nuclear Information System (INIS)

    Full text: Uranium-Plutonium mixed oxide (MOX) fuel for both thermal and fast reactors have been fabricated by Advanced Fuel Fabrication Facility (AFFF), Bhabha Atomic Research Centre, Tarapur, India. MOX fuel bundles fabricated by AFFF have been loaded in Boiling Water Reactors (BWRs) and Pressurised Heavy Water Reactors (PHWRs) and have been discharged after successful irradiation. An experimental fuel subassembly containing 37 MOX pins is being irradiated in Fast Breeder Test Reactor (FBTR) at Kalpakkam near Chennai and has seen a burn up of more than 80000 MWD/T. MOX fuel pins containing 44% Pu02 have been recently loaded as a part of the hybrid core of FBTR. AFFF has now taken up the manufacture of MOX fuel pins for the Prototype Fast Breeder Reactor (BHAVINI) coming up at Kalpakkam. The core consists of 181 sub assemblies containing 217 MOX fuel pins each. It is required to fabricate nearly 40,000 MOX fuel pins (3 meter long) for the first core. The Prototype Fast Breeder Reactor is designed with two different fissile enrichment zones to be loaded with MOX subassemblies with a nominal composition of 21% and 28% of PuO2. The fuel pellets of required composition are made using conventional powder metallurgy processes. The pellets are annular with an inner hole of 1.8mm diameter and outside diameter of 5.5mm. AFFF has developed the technology of making annular MOX fuel pellets for PFBR and optimized conditions of fabrication. Multistation rotary presses have been used for compaction of the pellets. The fuel pin consists of a MOX stack of 1000mm and axial blanket of deeply depleted uranium dioxide of length 300mm on either side. New techniques have been used at different stages of fabrication of the fuel pins namely pelletisation, welding and wire wrapping. Studies have been made to use laser welding technique for welding of endplugs. Automation has been introduced in a number of process steps in the fabrication line. A detailed quality control plan is prepared

  19. Fabrication of MOX Fuel elements for irradiation in Fast Breeder Test Reactor (FBTR)

    International Nuclear Information System (INIS)

    Advanced Fuel Fabrication Facility (AFFF), Bhabha Atomic Research Centre, Tarapur is fabricating Uranium - Plutonium Mixed Oxide Fuel (MOX) for different types of reactors. Recently MOX fuel pins for an experimental fuel subassembly of 37 pins has been fabricated for irradiation in Fast Breeder Test Reactor (FBTR) at Kalpakkam near Chennai. MOX fuel pins containing 44% PuO2 have also been also made for the hybrid core of FBTR. The experimental sub-assembly for irradiation testing in FBTR consisted of 37 short length Prototype Fast Breeder Reactor (PFBR) MOX fuel elements. The composition of the fuel was (0.71 U - 0.29 Pu) O2 with U233 O2 content of 53.5% of total UO2. Uranium enriched with U233 was used to simulate the heat flux of PFBR in FBTR neutron spectrum. MOX fuel pellets were made by powder metallurgy process consisting of pre-compaction, granulation, final compaction and sintering at high temperature. Initially U3233 O8 / U233 O3 powder was subjected to heat treatment. The pellets were sintered at reducing atmosphere at 1650oC for 4 hours to obtain acceptable quality pellets. Over sized pellets were centrelessly ground.without using a liquid coolant. During the fabrication of pins for experimental subassembly, technology was developed and conditions were optimized for making annular pellets, TIG welding of D9 tubes with SS 316 end plugs and wire wrapping. Quality control procedures and process control procedures at different stages of fabrication were developed. The hybrid core of FBTR consists of Mixed Carbide (MC) sub-assemblies containing (0.70 Pu - 0.30 U) C pellets and MOX fuel sub-assemblies containing (0.44 Pu - 0.56 U) O2. Studies were made to fabricate fuel containing higher percentage of Plutonium and the conditions were established. This paper describes the development of flowsheet for making annular MOX fuel pellets containing plutonium and U233, the technology for welding of D-9 clad tubes, wire wrapping and inspection. The paper also

  20. Clinch River Breeder Reactor

    International Nuclear Information System (INIS)

    Mr. Baron says the administration's effort to terminate the Clinch River Breeder Reactor (CRBR) project is symptomatic; they have also placed restrictions on fusion, coal, solar, and other areas of energy development in which technological advances are held back in order to force conservation. Because the breeder reactor, unlike solar and fusion energy, is both economically and technically feasible, a demonstration plant is needed. The contentions that the CRBR design is obsolete, that its proposed size is inappropriate, or that plutonium can be diverted for weapons proliferation are argued to be invalid. Failure to complete the CRBR will have both economic and national security repercussions

  1. Fast breeder reactor research

    International Nuclear Information System (INIS)

    Full text: The meeting was attended by 15 participants from seven countries and two international organizations. The Eighth Annual Meeting of the International Working Group on Fast Reactors (IWGFR) was attended by representatives from France, Fed. Rep. Germany, Italy, Japan, United Kingdom, Union of Soviet Socialist Republics and the United States of America - countries that have made significant progress in developing the technology and physics of sodium cooled fast reactors and have extensive national programmes in this field - as well as by representatives of the Commission of the European Communities and the IAEA. The design of fast-reactor power plants is a more difficult task than developing facilities with thermal reactors. Different reactor kinetics and dynamics, a hard neutron spectrum, larger integral doses of fuel and structural material irradiation, higher core temperatures, the use of an essentially novel coolant, and, as a result of all these factors, the additional reliability and safety requirements that are imposed on the planning and operation of sodium cooled fast reactors - all these factors pose problems that can be solved comprehensively only by countries with a high level of scientific and technical development. The exchange of experience between these countries and their combined efforts in solving the fundamental problems that arise in planning, constructing and operating fast reactors are promoting technical progress and reducing the relative expenditure required for various studies on developing and introducing commercial fast reactors. For this reason, the meeting concentrated on reviewing and discussing national fast reactor programmes. The situation with regard to planning, constructing and operating fast experimental and demonstration reactors in the countries concerned, the experience accumulated in operating them, the difficulties arising during operation and ways of over-coming them, the search for optimal designs for the power

  2. The development of breeder reactors in the US

    International Nuclear Information System (INIS)

    This article discusses the early history of breeder development in the US, the early history of the fast reactor in the US, changes during the Carter administration, and the development of LMFBR technology. Topics considered include the intermediate-energy plutonium breeder, the molten plutonium breeder, the aqueous homogeneous reactor, the molten-salt reactor, the liquid metal-fueled reactor, electronuclear breeding, the Experimental Breeder Reactor-I, the Experimental Breeder Reactor-II, the Enrico Fermi Reactor, a programmatic change to ceramic fuel, the South East Fast Oxide Reactor, the sodium void coefficient, the 1000-MWe studies of 1964, the 1000-MWe studies of 1967-1969, the FARET design, the Fast Flux Test Facility, the Clinch River Breeder Reactor (CRBR), the gas-cooled fast breeder, the light-water breeder, materials for cladding and duct walls, and reactor safety. It is pointed out that the Congress opposes the construction of the CRBR, while the Reagan administration strongly supports it

  3. Gamma-ray spectra of fast-breeder spent nuclear fuel from the BN-350 reactor

    International Nuclear Information System (INIS)

    Gamma-ray measurements of spent nuclear fuel (SNF) from a fast breeder reactor have been obtained with a High-Purity Germanium (HPGe) Detector. The HPGe measurements were performed inside a hot cell using an adjustable collimator to restrict the viewing angle of the HPGe to a small region of the SNF assembly. In addition Ion Chamber (IC) measurements were performed underwater using a lead shielded IC 15-cm in active length. We are going to present HPGe measurement results of the distribution of fission product and activation products along the assembly. We will also compare the gamma-ray profiles of the HPGe and IC measurements to those of the neutron profiles measured with a 3 He tube based neutron counter

  4. Potential of duplex fuel in prebreeder, breeder, and power reactor designs: tests and analyses (AWBA Development Program)

    International Nuclear Information System (INIS)

    Dual region fuel pellets, called duplex pellets, are comprised of an outer annular region of relatively high uranium fuel enrichment and a center pellet of fertile material with no enrichment. UO2 and ThO2 are the fissile and fertile materials of interest. Both prebreeders and breeders are discussed as are the performance advantages of duplex pellets over solid pellets in these two pressurized water reactor types. Advantages of duplex pellets for commercial reactor fuel rods are also discussed. Both irradiation test data and analytical results are used in comparisons. Manufacturing of duplex fuel is discussed

  5. Pulsed Nd-YAG laser welding of Prototype Fast Breeder Reactor fuel elements

    International Nuclear Information System (INIS)

    End plug welding of Prototype Fast Breeder Reactor (PFBR) fuel elements involves welding of fully Austenitic Stainless Steel (ASS) of grade D9 clad tube with 316M end plug. Pulsed Gas Tungsten Arc Welding (GTAW) is being used for the production of PFBR fuel elements at Advanced Fuel Fabrication Facility (AFFF). GTAW is an established process for end plug welding and hence adopted by many countries. GTAW has got certain limitations like heat input, arc gap sensitivity and certain sporadic defects like tungsten inclusion. Experiments have been carried out at AFFF to use Laser Beam Welding (LBW) technique as LBW offers a number of advantages over the former process. This report mainly deals with the optimization of laser parameters for welding of PFBR fuel elements. To facilitate pulsed Nd-YAG laser spot welding, parameters like peak power, pulse duration, pulse energy, frequency and defocusing of laser beam on to the work piece have been optimized. On the basis of penetration requirement laser welding parameters have been optimized. (author)

  6. Investigations on the mechanical interaction between fuel and cladding (FCMI) in fast breeder reactor fuel pins

    International Nuclear Information System (INIS)

    The relation between FCMI and plastic cladding distensions of Fast-Breeder pins with oxide as well as carbide fuel was analyzed theoretically and experimentally. This resulted in the possibility of plastic cladding straining caused by differential swelling of fuel and cladding material under stationary power conditions or differential thermal expansion at power changes. At stationary operating conditions the FCMI in oxide pins is limited by an irradiation-induced creep deformation into inner void volume and thus the fuel swelling pressure will never cause clad distensions worth mentioning. However, the cladding of carbide pins can be strained under stationary conditions because of the comparatively low fuel plastification under irradiation. Plastic straining of oxide pins may follow from differential thermal expansion at power changes. The amount of strain is primarily dependent upon magnitude and rate of the power increase, the starting conditions, and the clad material strength. The parameter dependence of the strains and the limiting conditions for their avoidance are reported. The model calculations are carried out by means of a special computer code which was developed following closely the results of irradiation experiments. It was proved experimentally that a considerably high geometrical swelling occurs after a power reduction until the fuel has come into contact with the cladding again. (orig.)

  7. Reactivity control capability of fuel-salt processing system in a molten-salt breeder reactor

    International Nuclear Information System (INIS)

    An evaluation is made of the reactivity control capability of the fuel processing system (FPS) in a molten-salt breeder reactor. The principal functions required of the FPS are: (a) Isolation of 233Pa from regions of high neutron flux during its decay to 233U, and (b) the removal of fission products from the system. The FPS can very usefully serve also to control the primary system reactivity by appropriately utilizing its function of extracting uranium and reconstituting the fuel contained in the salt. The principles of operation are quite similar to the chemical shim control system currently installed in PWR's whereby the core reactivity, affected by changes in the moderator temperature, fuel burnup and transient Xe, is adjusted by regulating the concentration of boric acid introducted in the moderator as neutron absorber. The present study examines the capability of the FPS to follow transient Xe as in PWR's, and proves that the FPS should effectively serve as a system for adjusting not only long-term changes in reactivity but also short-term transient variations without any accompanying difficulties foreseen in operation. (auth.)

  8. A study of parameters on marking of Prototype Fast Breeder Reactor fuel elements

    International Nuclear Information System (INIS)

    Prototype Fast Breeder Reactor Fuel (PFBR) elements are identified with a permanent unique marking. Identification of the fuel elements is very much necessary for traceability during initial fabrication as well as for post irradiation examination. Marking on fuel element has to be permanent and capable of being identified after irradiation. Laser marking is a relatively new method as compared to other marking technologies such as ink marking, mechanical engraving and electro chemical methods. It is used for the product identification and traceability during its service life. Laser marking has many advantages compared to other conventional marking. In laser marking process, mark quality is a very important factor, which depends on so many variables like input current, pulse frequency, marking speed and number of passes. The influence of the pulse frequency and the speed of travel of the laser beam on the mark depth and width have been studied in this paper. An optical microscope, scanning electron microscope were used to measure the effects of pulse frequency on the mark depth and width. It has been found that the mark depth and width depend on the interaction process of the laser beam and the material, which was influenced by the pulse frequency. Micro hardness testing is carried out to report Heat Affected Zone (HAZ) variation with parameters. Marking speed and input current selected for suitable depth and width were mentioned in the present study. (author)

  9. Fabrication and loading of fuel rods for the Light Water Breeder Reactor (LWBR Development Program)

    International Nuclear Information System (INIS)

    The fabrication and inspection operations used for the manufacture of approximately 24,000 fuel rods for the Light Water Breeder Reactor are described in detail. This report also describes the development work to establish the fabrication procedures and investigations undertaken to solve problems encountered during manufacturing. The approximately 10 foot long LWBR fuel rods were made in four outside diameters ranging from 0.306 inch (seed) to 0.832 inch (reflector). Each rod was fabricated by sealing cylindrical oxide fuel pellets (ThO2-U233O2), into Zircaloy seamless tubes by welding Zircaloy enclosures at the ends. The special inspections performed to assure a high quality product meeting all design requirements are described. These inspections included weld radiography and ultrasonic inspection, in-motion radiography to evaluate internal dimensions and pellet integrity, helium leak testing, corrosion testing, and detection of surface contamination. The facilities designed and built for this fabrication effort are described and the resultant manufacturing yields are presented. 13 refs., 42 figs., 20 tabs

  10. Status of fast breeder reactors and associated fuel cycle in India

    International Nuclear Information System (INIS)

    Full text: India is the largest democracy with the current population of about 1.05 billion and is on a road to rapid growth in economy. An impressive average domestic product (GDP) growth rate of about 8 % per year has been achieved in 2006-07 and it is targeted to touch 10 % per year for the next 10 years. Towards realizing this targeted growth, development activities are planned based on well-conceived road map and clear vision. Like elsewhere, the energy and electricity growth in India are also closely linked to growth in economy. Indices of socio-economic development like literacy, longevity, GDP and human development are directly dependent upon the per capita energy consumption of a country. India is aiming to reach at least per capita energy consumption equal to the present world average (2200 kWh/a) by 2030 from the current value of (660 kWh/a). The current installed capacity of ∼138 GW(e) needs to be increased to about 600 GWe by 2030 assuming the population of about 1.4 billion. Energy strategists in the country have realized the importance of judicious mix of energy resources to meet this challenge. A large share of nuclear energy is an inevitable choice in this judicious energy mix from resources, sustainability and environment considerations. The nuclear is expected to contribute about 63 GWe by 2030, which will be steadily increased to 275 GWe by 2052, against the total projected capacity of 1344 GWe. The three stage visionary programme of India envisages Water Reactors (first stage), Fast Breeders with high breeding (second stage) and Thorium based Reactors as third stage. Closed fuel cycle in all stages is an essential ingredient. The success of each stage depends upon expeditious maturity of the earlier stage as India has limited indigenous resources of uranium, but vast resources of thorium. India ranks high in nuclear technology scale with strong R and D, high quality human resources, sound infrastructure, unwavering Government support and

  11. Fast breeder reactors an engineering introduction

    CERN Document Server

    Judd, A M

    1981-01-01

    Fast Breeder Reactors: An Engineering Introduction is an introductory text to fast breeder reactors and covers topics ranging from reactor physics and design to engineering and safety considerations. Reactor fuels, coolant circuits, steam plants, and control systems are also discussed. This book is comprised of five chapters and opens with a brief summary of the history of fast reactors, with emphasis on international and the prospect of making accessible enormous reserves of energy. The next chapter deals with the physics of fast reactors and considers calculation methods, flux distribution,

  12. Considerations of the effects of azimuthal fuel motion in a fast breeder reactor accident

    International Nuclear Information System (INIS)

    A sizeable reactivity feedback can result from material movement in a large liquid metal fast breeder reactor (LMFBR). Previous investigations considered mainly fuel slumping under gravity and outward radial motion. Very little work has been done on azimuthal motion. Furthermore, studies of the effects of material motion either use point kinetics or very expensive space-time differencing solutions. This work was undertaken to develop an intermediate approach between point kinetics and the full space-time finite difference solutions. The approach was then applied to sample problems with azimuthal and radial material motion under possible accident conditions. Specifically, the objectives of the work were to: (a) develop a technique for treating space-time neutron kinetics during postulated accident conditions which include material motion; (b) apply the technique to sample problems. The technique was developed based on the use of the finite element method (FEM) for the spatial differencing of the multigroup, time dependent diffusion equation. The FEM was chosen for three reasons. First, the FEM gives good accuracy with relatively fewer unknowns than the finite difference method. Second, the FEM is very flexible in setting up a mesh for the geometry of concern. Last, the FEM could handle the spatial differencing of a mesh which became distorted as the material in the reactor moved. This material motion was handled by specifying the FEM mesh nodes as a function of time and periodically updating the spatial matrices. Finally, the method used to solve the time dependence was Gear's variable order predictor corrector scheme

  13. Fabrication of MOX Fuel elements for irradiation in Fast Breeder Test Reactor (FBTR)

    International Nuclear Information System (INIS)

    Advanced Fuel Fabrication Facility (AFFF), Bhabha Atomic Research Centre, Tarapur is fabricating Uranium - Plutonium Mixed Oxide Fuel (MOX) for different types of reactors. Recently an experimental fuel subassembly of 37 pins has been fabricated for irradiation in Fast Breeder Test Reactor at Kalpakkam near Chennai. MOX fuel pins containing 45% PUO2 have also been also made for the hybrid core of FBTR. The experimental sub-assembly for irradiation testing in FBTR consists of 37 short length PFBR MOX fuel elements. The composition of the fuel was (0.71 U - 0.29 Pu) O2 with U233 O2 content of 53.5% of total UO 2. Uranium enriched with U233 was used to stimulate the heat flux of PFBR in FBTR neutron spectrum. MOX fuel pellets were made by powder metallurgy process consisting of pre-compaction, granulation, final compaction and sintering at high temperature. Initially U3233 O8 / U233 O3 powder was subjected to heat treatment. MOX powder were mixed, milled, pre-compacted and granulated. The final compaction was done using a multistation rotary press with suitable tooling for making annular MOX pellets. The technology for making annular pellets was developed for this purpose. The pellets were sintered at reducing atmosphere at 1650 deg. C for 4 hours to obtain acceptable quality pellets. Over sized pellets were centrelessly ground without using a liquid coolant. The acceptable pellets were degassed before encapsulation. MOX fuel stack, UO2 insulation pellets, plenum spring and spring support were loaded in bottom endplug welded clad tube. The end plug welding was carried out by TIG welding technique. The welded elements after inspection were wire wrapped. During the fabrication of pins for experimental subassembly, technology was developed and conditions were optimized for making annular pellets, TIG welding of D9 tubes with SS 316 end plugs and wire wrapping. Quality control procedures and process control procedures at different stages of fabrication were developed. The

  14. Future designs of breeder reactors (Europe, USA)

    International Nuclear Information System (INIS)

    Sodium-cooled reactors with a fast neutron core today are the only fission reactors that offer the reactor physics required for the breeding process and the complete conversion of U-238 or Th-232 into fissile fuel. There are currently five prototype breeder reactors in operation in England, France, and the USSR. The trends observable in development work aim at reducing capital cost, enhancing and improving passive shutdown performance, and simplifying the fuel cycle. (orig.)

  15. Optimal measurement uncertainties for materials accounting in a fast breeder reactor spent-fuel reprocessing plant

    International Nuclear Information System (INIS)

    Optimization techniques are used to calculate measurement uncertainties for materials accountability instruments in a fast breeder reactor spent-fuel reprocessing plant. Optimal measurement uncertainties are calculated so that performance goals for detecting materials loss are achieved while minimizing the total instrument development cost. Improved materials accounting in the chemical separations process (111 kg Pu/day) to meet 8-kg plutonium abrupt (1 day) and 40-kg plutonium protracted (6 months) loss-detection goals requires: process tank volume and concentration measurements having precisions less than or equal to 1%; accountability and plutonium sample tank volume measurements having precisions less than or equal to 0.3%, short-term correlated errors less than or equal to 0.04%, and long-term correlated errors less than or equal to 0.04%; and accountability and plutonium sample tank concentration measurements having precisions less than or equal to 0.4%, short-term correlated errors less than or equal to 0.1%, and long-term correlated errors less than or equal to 0.05%

  16. Design of fuel fabrication plant of Fast Reactor Fuel Cycle Facility for reload requirement of Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    India's economic growth is on a fast growth track. The energy demand is expected to grow rapidly in the coming decades. The growth in population and economy is creating huge demand for energy which has to be met with environmentally benign technologies. Nuclear energy is best suited to meet this demand in a sustainable manner without causing undue environmental impact. Fast reactors are expected to be major contributors in sufficing this demand to a great extent. As an effort to achieve the objective, a Prototype Fast Breeder Reactor is being constructed at Kalpakkam. This paper also highlights the design features of FFP, unit operations, scheme of automation, branched layout of glove box train, shielding arrangement on glove boxes, accident consequence analysis etc.

  17. Post-irradiation examination of mixed (Pu, U)C fuels irradiated in the fast breeder reactor

    International Nuclear Information System (INIS)

    The Fast Breeder Test Reactor (FBTR) at Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, India, using mixed (Pu,U)C fuel has completed eleven years of operation. One fuel subassembly which has seen more than 25,000 MWd/t burn-up has been discharged from this reactor taken up for post-irradiation examination. The PIE carried out on this fuel subassembly has established that the fuel has performed satisfactorily and it is capable of being taken to higher levels of burn-up and linear heat ratings. The facilities available for PIE of advanced fuels and the PIE work carried out are discussed in detail in this paper. (author)

  18. Manufacturing experience for mixed uranium-plutonium carbide fuels for fast breeder test reactor

    International Nuclear Information System (INIS)

    The plutonium rich mixed uranium-plutonium carbide pellets of two compositions, namely (U0.3Pu0.70)C (MK-I) and (U0.45Pu0.55)C (MK-II), are used as the fuel for the Indian Fast Breeder Test Reactor (FBTR) at Kalpakkam. These fuels were developed and are being fabricated and characterized at Bhabha Atomic Research Centre (BARC) and have performed very well with peak burn-up exceeding 155GWd/t. This achievement has been possible through a combination of stringent fuel specifications, quality control during fabrication and inputs obtained from the detailed post irradiation examination of fuel at different stages combined with the modeling of the behaviour of the fuel clad and wrapper materials. The high burn-up and short cooled fuel has also been reprocessed successfully in the reprocessing facility at IGCAR. The fissile material (Pu) recovered from reprocessing has now been used for fabrication of fresh mixed carbide fuel which will be loaded in FBTR in the next reload schedule. Closing the carbide fuel cycle is an important milestone in the fast reactor fuel cycle. Bhabha Atomic Research Centre, Trombay developed the fabrication flow sheet for MK-I and MK-II carbide fuels for FBTR. Since carbide fuel is pyrophoric and susceptible to hydrolysis, the fabrication has to be carried out in high purity nitrogen cover gas in leak tight glove boxes. Moreover, adequate shielding is provided to minimize the personnel exposure. The carbide fuel are made using powder metallurgy route with UO2, PuO2 and graphite as the staring material. The homogeneously mixed oxide and graphite powders are compacted into small tablets at low pressure in order to have handling strength and intimate contact between oxide and graphite particles, and to have sufficient porosities for the easy removal of carbon monoxide. The vacuum and temperature for carbothermic reduction are controlled in order to minimize plutonium losses by vaporization and also to have oxygen, nitrogen, carbon, higher carbide

  19. Multiple recycling of fuel in prototype fast breeder reactor in a closed fuel cycle with pressurized heavy-water reactor external feed

    Indian Academy of Sciences (India)

    G Pandikumar; A John Arul; P Puthiyavinayagam; P Chellapandi

    2015-10-01

    A fast breeder reactor (FBR) closed fuel cycle involves recycling of the discharged fuel, after reprocessing and refabrication, in order to utilize the unburnt fuel and the bred fissile material. Our previous study in this regard for the prototype fast breeder reactor (PFBR) indicated the possibility of multiple recycling with self-sufficiency. It was found that the change in Pu composition becomes negligible (less than 1%) after a few cycles. The core-1 Pu increases by 3% from the beginning of cycle-0 to that of recycle-1, the Pu increase from the beginning of the 9th cycle to that of the 10th by only 0.3%. In this work, the possibility of multiple recycling of PFBR fuel with external plutonium feed from pressurized heavy-water reactor (PHWR) is examined. Modified in-core cooling and reprocessing periods are considered. The impact of multiple recycling on PFBR core physics parameters due to the changes in the fuel composition has been brought out. Instead of separate recovery considered for the core and axial blankets in the earlier studies, combined fuel recovery is considered in this study. With these modifications and also with PHWR Pu as external feed, the study on PFBR fuel recycling is repeated. It is observed that the core-1 initial Pu inventory increases by 3.5% from cycle-0 to that of recycle-1, the Pu increase from the beginning of the 9th cycle to that of the 10th is only 0.35%. A comparison of the studies done with different external plutonium options viz., PHWR and PFBR radial blanket has also been made.

  20. Gas-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Almost all the R D works of gas-cooled fast breeder reactor in the world were terminated at the end of the year 1980. In order to show that the R D termination was not due to technical difficulties of the reactor itself, the present paper describes the reactor plant concept, reactor performances, safety, economics and fuel cycle characteristics of the reactor, and also describes the reactor technologies developed so far, technological problems remained to be solved and planned development schedules of the reactor. (author)

  1. Summary of estimated doses and risks resulting from routine radionuclide releases from fast breeder reactor fuel cycle facilities

    International Nuclear Information System (INIS)

    A project is underway at Oak Ridge National Laboratory to assess the human health and environment effects associated with operation of Liquid Metal Fast Breeder Reactor fuel cycle. In this first phase of the work, emphasis was focused on routine radionuclide releases from reactor and reprocessing facilities. For this study, sites for fifty 1-GW(e) capacity reactors and three reprocessing plants were selected to develop scenarios representative of US power requirements. For both the reactor and reprocessing facility siting schemes selected, relatively small impacts were calculated for locality-specific populations residing within 100 km. Also, the results of these analyses are being used in the identification of research priorities. 13 refs., 2 figs., 3 tabs

  2. Estimated doses and risks resulting from routine radionuclide releases from fast breeder reactor fuel cycle facilities: a summary

    International Nuclear Information System (INIS)

    A project has been carried out at Oak Ridge National Laboratory to assess the human health and environmental effects associated with the operation of a liquid-metal fast breeder reactor fuel cycle. In this first phase of the work, emphasis was on routine radionuclide releases from reactor and reprocessing facilities. Sites for 51 1-GW(e) capacity reactors and 3 reprocessing plants were selected to develop scenarios representative of US power requirements. For both the reactor and the reprocessing facility siting schemes selected, relatively small impacts were calculated for locality-specific populations residing within 100 km. Also the results of these analyses are being used in the identification of research priorities

  3. Theoretical and experimental studies of non-linear structural dynamics of fast breeder reactor fuel elements

    International Nuclear Information System (INIS)

    Descriptions are presented of theoretical and experimental studies of the deformation behaviour of fast-breeder fuel elements as a consequence of extreme impulsive stresses produced by an incident. The starting point for the studies is the assumption that local disturbances in a fuel element have resulted in a thermal interaction between fuel and sodium and in a corresponding increase in pressure. On the basis of the current state of knowledge, the possibility cannot be ruled out that this pressure build-up may lead to the bursting of the fuel-element wrapper, to the propagation of pressure in the core, and to coherent structural movements and deformations. A physical model is established for the calculation of the dynamic response of elastic-plastic beam systems, and the differential equations of p motion for the discrete equivalent system are derived with the aid of D'Alembert's principle. On this basis and with the aid of a semi-empirical pin-bundle model, an appropriate computer program allows a static and dynamic analysis to be obtained for a complete fuel element. In the experimental part of the study, a description is given of static and impulsive loading tests on 1:1 SNR-like fuel-element models. Making use of measured impact forces and of known material characteristics, it was possible to a large extent for the experiments to be reproduced by calculations. In agreement with existing experience from explosion experiments on 1:1 core models, the results (of relevance for fast-breeder safety and in particular the SNR-300) show that only local limited deformations occur and that the compact fuel-element and core structure constitutes an effective inherent barrier in the presence of extreme incident stresses. (author)

  4. Fast breeder reactor

    International Nuclear Information System (INIS)

    This paper outlined the present status of FBR development in six countries and reviewed Japanese activities on FBR development. Joyo experimental FBR has accumulated a lot of technical data including irradiation tests of advanced fuels and was now long shut down due to the partial obstruction of rotating plug movement. Monju prototype FBR reactor experienced a sodium leakage in its secondary heat transfer system during performance tests in December 1995 and had been shut down until May 2010. Feasibility study on commercialized FBR cycle system ended in March 2006 and proposed the concept of commercialized FBR cycle technologies. In order to plan a demonstration reactor, research and development of innovative technologies are conducted as the FaCT (Fast Reactor Cycle Technology Development) Project. In connection with the results of this research and development, a 5-party council of Japan was established to discuss processes of demonstration and commercialization of FBR cycle systems in Japan. Joint efforts were made for a demonstration reactor to be committed in 2015, in addition to start operation around 2025 aiming at the commercialization of FBR before 2050. (T. Tanaka)

  5. An investigation of nuclear physics characteristics of fast breeder reactors (LMFBR and GCFBR) with various fuel cycles

    International Nuclear Information System (INIS)

    The primary emphasis on the study has been placed on comparing neutronic characteristics, e.g. fissile inventory, breeding and safety, of fast breeder reactors with uranium-plutonium and thorium-uranium fuel cycles. The study was performed using identical calculation methods and consistent data basis. As the reference fast breeder reactor, two different types of 1,200 MWe PuO2-UO2 fuelled fast reactors were chosen, which are sodium-cooled fast breeder reactor (LMFBR) and helium-cooled fast breeder reactor (GCFBR). The following four fuel utilisation models were investigated for each of LMFBR and GCFBR. (1) PuO2-UO2 core, and UO2 axial and radial blankets, (2) PuO2-UO2 core, UO2 axial blanket and ThO2 radial blanket, (3) 233UO2-UO2 core, and ThO2 axial and radial blankets, (4) 233UO2-ThO2 core, and ThO2 axial and radial blankets. The main results obtained are summarised as follows: (1) Pu fuelled LMFBR provides sufficiently high breeding gain, but has unfavourable characteristics of considerably large positive sodium-void reactivity effect. (2) U-233 fuelled LMFBR provides the favourable characteristics of negative sodium-void reactivity effect, but provides either negative or very low breeding gain. (3) Pu fuelled GCFBR has the desirable characteristics from the viewpoints investigated in the study, i.e. relatively low fissile inventory, very large breeding gain, sufficiently negative Doppler reactivity effect and negative steam ingress reactivity effect. (4) Use of U-233 in the core of GCFBR is not preferable, because of substantially low breeding gain and terribly large positive steam ingress reactivity effect. (5) Use of ThO2 in the core of LMFBR and GCFBR instead of UO2 leads to increase of fissile inventory and decrease of breeding gain. (6) Use of ThO2 in the blanket of LMFBR and GCFBR instead of UO2 does not give any significant influence on the neutronic characteristics

  6. Liquid metal cooled fast breeder nuclear reactor

    International Nuclear Information System (INIS)

    A liquid metal cooled fast breeder nuclear reactor has a core comprising a plurality of fuel assemblies supported on a diagrid and submerged in a pool of liquid metal coolant within a containment vessel, the diagrid being of triple component construction and formed of a short cylindrical plenum mounted on a conical undershell and loosely embraced by a fuel store carrier. The plenum merely distributes coolant through the fuel assemblies, the load of the assemblies being carried by the undershell by means of struts which penetrate the plenum. The reactor core, fuel store carrier and undershell provide secondary containment for the plenum. (UK)

  7. EPRI Asilomar papers: on the possibility of advanced fuel fusion reactors, fusion-fission hybrid breeders, small fusion power reactors, Asilomar, California, December 15--17, 1976

    International Nuclear Information System (INIS)

    An EPRI Ad Hoc Panel met in Asilomar, California for a three day general discussion of topics of particular interest to utility representatives. The three main topics considered were: (1) the possibility of advanced fuel fusion reactors, (2) fusion-fission hybrid breeders, and (3) small fusion power reactors. The report describes the ideas that evolved on these three topics. An example of a ''neutron less'' fusion reactor using the p-11B fuel cycle is described along with the critical questions that need to be addressed. The importance to the utility industry of using fusion neutrons to breed fission fuel for LWRs is outlined and directions for future EPRI research on fusion-fission systems are recommended. The desirability of small fusion power reactors to enable the early commercialization of fusion and for satisfying users' needs is discussed. Areas for possible EPRI research to help achieve this goal are presented

  8. Computerized operating procedures for shearing and dissolution of segments from LWBR [Light Water Breeder Reactor] fuel rods

    International Nuclear Information System (INIS)

    This report presents two detailed computerized operating procedures developed to assist and control the shearing and dissolution of irradiated fuel rods. The procedures were employed in the destructive analysis of end-of-life fuel rods from the Light Water Breeder Reactor (LWBR) that was designed by the Westinghouse Electric Corporation Bettis Atomic Power Laboratory. Seventeen entire fuel rods from the end-of-life core of the LWBR were sheared into 169 precisely characterized segments, and more than 150 of these segments were dissolved during execution of the LWBR Proof-of-Breeding (LWBR-POB) Analytical Support Project at Argonne National Laboratory. The procedures illustrate our approaches to process monitoring, data reduction, and quality assurance during the LWBR-POB work

  9. Computerized operating procedures for shearing and dissolution of segments from LWBR (Light Water Breeder Reactor) fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Osudar, J.; Deeken, P.G.; Graczyk, D.G.; Fagan, J.E.; Martino, F.J.; Parks, J.E.; Levitz, N.M.; Kessie, R.W.; Leddin, J.M.

    1987-05-01

    This report presents two detailed computerized operating procedures developed to assist and control the shearing and dissolution of irradiated fuel rods. The procedures were employed in the destructive analysis of end-of-life fuel rods from the Light Water Breeder Reactor (LWBR) that was designed by the Westinghouse Electric Corporation Bettis Atomic Power Laboratory. Seventeen entire fuel rods from the end-of-life core of the LWBR were sheared into 169 precisely characterized segments, and more than 150 of these segments were dissolved during execution of the LWBR Proof-of-Breeding (LWBR-POB) Analytical Support Project at Argonne National Laboratory. The procedures illustrate our approaches to process monitoring, data reduction, and quality assurance during the LWBR-POB work.

  10. Concept and Development Status of Fast Breeder Reactor Fuels in the FaCT Project

    International Nuclear Information System (INIS)

    A conceptual design study and related R&D on the JSFR (Japan sodium cooled fast reactor) with mixed oxide (MOX) fuels, advanced aqueous reprocessing and simplified pelletizing fuel fabrication as a promising concept have been implemented in the fast reactor cycle technology development (FaCT) project. The fuel concept is being established in the FaCT project to improve economic potential in the fuel cycle and to enhance safety characteristics. Ferritic core materials and large diameter fuel pins with annular pellets will be adopted in the high burnup fuel. An inner duct will be equipped in the fuel subassembly to mitigate the core disruptive event. To actualize the concept, key technologies to be developed are oxide dispersion strengthened (ODS) ferritic steel with high temperature mechanical strength for fuel pin claddings and the simplified pelletizing fuel fabrication system, including a microwave de-nitration of plutonium enrichment adjusted solution and die wall lubrication. In the present status of the project, many basic technical findings of ODS ferritic steel have been obtained in the field of powder metallurgy, mechanical properties and irradiation characteristics. The application potential of the simplified pelletizing method has been confirmed. Furthermore, the properties of MOX fuel bearing minor actinides (MAs), including melting point and thermal conductivity, have been systematically measured to develop the MA-bearing MOX fuel with the aim of reducing the amount and the toxicity of radioactive wastes. The design technology of the MA-bearing MOX fuel with annular pellets has been also studied. (author)

  11. Fast breeder fuel cycle

    International Nuclear Information System (INIS)

    Basic elements of the ex-reactor part of the fuel cycle (reprocessing, fabrication, waste handling and transportation) are described. Possible technical and proliferation measures are evaluated, including current methods of accountability, surveillance and protection. The reference oxide based cycle and advanced cycles based on carbide and metallic fuels are considered utilizing conventional processes; advanced nonaqueous reprocessing is also considered. This contribution provides a comprehensive data base for evaluation of proliferation risks

  12. The passive nondestructive assay of the plutonium content of spent-fuel assemblies from the BN-350 fast-breeder reactor in the city of Aqtau, Kazakhstan

    CERN Document Server

    Lestone, J P; Rennie, J A; Sprinkle, J K; Staples, P; Grimm, K N; Hill, R N; Cherradi, I; Islam, N; Koulikov, J; Starovich, Z

    2002-01-01

    The International Atomic Energy Agency is presently interested in developing equipment and techniques to measure the plutonium content of breeder reactor spent-fuel assemblies located in storage ponds before they are relocated to more secure facilities. We present the first quantitative nondestructive assay of the plutonium content of fast-breeder reactor spent-fuel assemblies while still underwater in their facility storage pond. We have calibrated and installed an underwater neutron coincidence counter (Spent Fuel Coincidence Counter (SFCC)) in the BN-350 reactor spent-fuel pond in Aqtau, Kazakhstan. A procedure has been developed to convert singles and doubles (coincidence) neutron rates observed by the SFCC into the total plutonium content of a given BN-350 spent-fuel assembly. The plutonium content has been successfully determined for spent-fuel assemblies with a contact radiation level as high as approx 10 sup 5 Rads/h. Using limited facility information and multiple measurements along the length of spe...

  13. Advanced fuel for fast breeder reactors: Fabrication and properties and their optimization

    International Nuclear Information System (INIS)

    The present design for FBR fuel rods includes usually MOX fuel pellets cladded into stainless steel tubes, together with UO2 axial blanket and stainless steel hexagonal wrappers. Mixed carbide, nitride and metallic fuels have been tested as alternative fuels in test reactors. Among others, the objectives to develop these alternative fuels are to gain a high breeding ratio, short doubling time and high linear ratings. Fuel rod and assembly designers are now concentrating on finding the combination of optimized fuel, cladding and wrapper materials which could result in improvement of fuel operational reliability under high burnups and load-follow mode of operation. The purpose of the meeting was to review the experience of advanced FBR fuel fabrication technology, its properties before, under and after irradiation, peculiarities of the back-end of the nuclear fuel cycle, and to outline future trends. As a result of the panel discussion, the recommendations on future Agency activities in the area of advanced FBR fuels were developed. A separate abstract was prepared for each of the 10 presentations of this meeting. Refs, figs and tabs

  14. Development and performance of fuel elements for sodium-cooled breeder reactors in Germany

    International Nuclear Information System (INIS)

    The first sodium-cooled reactor commissioned in Germany, KNK, serves now as test facility for plutonium bearing oxide fuel elements. The target is to provide reliable fuel for the SNR-300 project (Kalkar Nuclear Power Plant). The long-range target is fuel for burnups above 100,000 MW d/t, which moreover can easily be fabricated and reprocessed. As in the U.K., the line of grid-spaced bundles is favorised, being promising as regards the possibility of replacement of a defected pin and reinsertion of the bundle. (orig.)

  15. Improved structural materials for fast breeder reactors

    International Nuclear Information System (INIS)

    Electricity plays a crucial role in the economic development of our country. Coal is the primary fuel for generation of electricity in India as in many other countries. In India, generation of power by nuclear reactors is very important because of (i) availability of large thorium resource, (ii) constraints on setting up of fossil fuel based power plants and (iii) the negligibly small green house gas emissions by nuclear energy. The nuclear programme of the country is being implemented in three stages: (i) pressurized heavy water reactors of the CANDU type, (ii) sodium-cooled fast reactors and (iii) thorium-based reactors. Sodium-cooled fast reactor (SFR) technology is envisioned to make use of the large thorium reserves available. India has undertaken and made rapid strides in developing SFR technology and building of fast reactors for energy generation. A Fast Breeder Test Reactor (FBTR) of 40 MWt is operating successfully for over 25 years at Indira Gandhi Centre for Atomic Research. Based on the design, construction and operational experience, a 500 MWe Prototype Fast Breeder Reactor (PFBR) has been designed indigenously and is in an advanced stage of construction. Its design is being further optimised for enhanced economy with respect to cost of electricity production, for use in commercial reactors. Currently, several R and D programmes are under implementation for the development of new materials required for improved economy of commercial fast reactors

  16. The Role of Energetic Mixed-Oxide-Fuel-Sodium Thermal Interactions in Liquid Metal Fast Breeder Reactor Safety

    International Nuclear Information System (INIS)

    Recent efforts dealing with the consequence assessment of low-probability core-disruptive accidents (CDAs) in liquid-metal fast breeder reactors (LMFBRs) suggest that unrealistic physical processes must be postulated in order to achieve energetic prompt burst conditions leading to a true hydrodynamic disassembly of the reactor core. Such calculations are, however, being used in the licensing process in order to provide an estimate of safety margins provided by a given design. Figure 1 illustrates calculations for the Fast Flux Test Facility (FFTF) and the Clinch River Breeder Reactor (CRBR), where the prompt critical excursion and associated ramp rates are induced by postulating various amounts and rates of collapsing fuel in a largely molten core (recriticality accident), and the mode of energy release considered is the expansion of fuel vapor resulting in sodium-slug impact on the reactor vessel head. The VENUS-II code is used to calculate the disassembly motion and power histories during disassembly Elementary thermodynamic calculations provide the source term based upon expansion of the fuel from an initial temperature distribution specified by VENUS calculations, and the REXCO series of codes provide a hydrodynamic calculation of the pressure propagation coupled with an analysis of the structural response of the important system components. The work potential resulting from fuel collapse and hydrodynamic disassembly is very sensitive to small variations in the ramp rate. Since material motions associated with postulated conditions leading to energetic prompt critical excursions cannot be described with sufficient accuracy to provide reasonable bounds on ramp rates, an adequate margin of safety with current design is difficult to claim if these conditions cannot be ruled out. This implies that in addition to coherent gravity collapse, the possibility of pressure-driven (fuel-coolant interaction) collapse must be considered. Furthermore, the work potential

  17. International cooperation on breeder reactors

    International Nuclear Information System (INIS)

    In March 1977, as the result of discussions which began in the fall of 1976, the Rockefeller Foundation requested International Energy Associates Limited (IEAL) to undertake a study of the role of international cooperation in the development and application of the breeder reactor. While there had been considerable international exchange in the development of breeder technology, the existence of at least seven major national breeder development programs raised a prima facie issue of the adequacy of international cooperation. The final product of the study was to be the identification of options for international cooperation which merited further consideration and which might become the subject of subsequent, more detailed analysis. During the course of the study, modifications in U.S. breeder policy led to an expansion of the analysis to embrace the pros and cons of the major breeder-related policy issues, as well as the respective views of national governments on those issues. The resulting examination of views and patterns of international collaboration emphasizes what was implicit from the outset: Options for international cooperation cannot be fashioned independently of national objectives, policies and programs. Moreover, while similarity of views can stimulate cooperation, this cannot of itself provide compelling justification for cooperative undertakings. Such undertakings are influenced by an array of other national factors, including technological development, industrial infrastructure, economic strength, existing international ties, and historic experience

  18. History and evolution of the breeder reactor

    International Nuclear Information System (INIS)

    The concept of the breeder reactor is almost as old as the idea of the nuclear reactor itself. From the very first years following the discovery of nuclear fission, scientists and technicians tried to turn mankind's eternal dream into reality; that is, enjoy an abundant source of energy without using up our raw material reserves. Nuclear energy offered several solutions to realize this dream. One of them, fusion, seemed out of our grasp in the near future. But fission of 235U was possible, and the Manhattan Project soon furnished ample proof of this theory. However, everyone working in this field was conscious of the fact that thermal neutron reactors make very inefficient use of the energy potential contained in natural uranium. The solution was to use in a core sufficiently rich in fissile matter, the excess neutrons to convert the 238U, so poorly used by other types of reactors, into fissile 239Pu. Regeneration, or 'breeding' of fuel, can multiply the energy drawn from a ton of uranium by a factor of 50 to 100. This would enable us to ward off the specter of an energy shortage and the rapid depletion of uranium mines. As early as 1945 in Los Alamos, Enrico Fermi stated: 'The country which first develops a breeder reactor will have a great competitive edge in atomic energy.' The development of the breeder reactor in the USA and around the world is discussed

  19. Core loading pattern optimization of thorium fueled heavy water breeder reactor using genetic algorithm

    International Nuclear Information System (INIS)

    In this work genetic algorithm was proposed to solve fuel loading pattern optimization problem in thorium fueled heavy water reactor. The objective function of optimization was to maximize the conversion ratio and minimize power peaking factor. Those objectives were simultaneously optimized using non-dominated Pareto-based population ranking optimal method. Members of non-dominated population were assigned selection probabilities based on their rankings in a manner similar to Baker's single criterion ranking selection procedure. A selected non-dominated member was bred through simple mutation or one-point crossover process to produce a new member. The genetic algorithm program was developed in FORTRAN 90 while neutronic calculation and analysis was done by COREBN code, a module of core burn-up calculation for SRAC. (authors)

  20. Thorium utilization in fast breeder reactors and in cross-progeny fuel cycles

    International Nuclear Information System (INIS)

    Thorium fuel cycles have to be closed since the benefit is obtained only when the 233U is used. India is the only country in the world, which has extensive facilities for reprocessing of irradiated Uranium and Thorium-based fuels, thermal reactors moderated by light and heavy water and 500 MWe LMFBRs. The cross-progeny fuel cycles would be a natural vision to pursue for India. This paper was written in 1982 and presented at the U.S. Japan Seminar on Thorium fuel cycle held in October 1982. The calculations performed and the results quoted in this paper are of that vintage. However, the cross section data for Th and other materials has not changed significantly since that time. The same holds for the methodologies in computer codes, diffusion theory and the other methodologies employed in this paper, versus those in computer codes currently in use. This paper is being submitted to remind the community that with the introduction of GEN IV LMFBRs, other possibilities for thorium utilization could spring forth and should be studied further and in more depth

  1. Review of the SIMMER-II analyses of liquid-metal-cooled fast breeder reactor core-disruptive accident fuel escape

    International Nuclear Information System (INIS)

    Early fuel removal from the active core of a liquid-metal-cooled fast breeder reactor undergoing a core-disruptive accident may reduce the potential for large energetics resulting from recriticalities. This paper presents a review of analyses with the SIMMER-II computer program of the effectiveness of possible fuel escape paths. Where possible, how SIMMER-II compares with or is validated against experiments that simulated the escape paths also is discussed

  2. Prototype fast breeder reactor main options

    International Nuclear Information System (INIS)

    Fast reactor programme gets importance in the Indian energy market because of continuous growing demand of electricity and resources limited to only coal and FBR. India started its fast reactor programme with the construction of 40 MWt Fast Breeder Test Reactor (FBTR). The reactor attained its first criticality in October 1985. The reactor power will be raised to 40 MWt in near future. As a logical follow-up of FBTR, it was decided to build a prototype fast breeder reactor, PFBR. Considering significant effects of capital cost and construction period on economy, systematic efforts are made to reduce the same. The number of primary and secondary sodium loops and components have been reduced. Sodium coolant, pool type concept, oxide fuel, 20% CW D9, SS 316 LN and modified 9Cr-1Mo steel (T91) materials have been selected for PFBR. Based on the operating experience, the integrity of the high temperature components including fuel and cost optimization aspects, the plant temperatures are recommended. Steam temperature of 763 K at 16.6 MPa and a single TG of 500 MWe gross output have been decided. PFBR will be located at Kalpakkam site on the coast of Bay of Bengal. The plant life is designed for 30 y and 75% load factor. In this paper the justifications for the main options chosen are given in brief. (author). 2 figs, 2 tabs

  3. Development of variable-width ribbon heating elements for liquid-metal and gas-cooled fast breeder reactor fuel-pin simulators

    International Nuclear Information System (INIS)

    Variable-width ribbon heating elements that provide a chopped-cosine variable heat flux profile have been fabricated for fuel pin simulators used in test loops by the Breeder Reactor Program Thermal-Hydraulic Out-of-Reactor Safety test facility and the Gas-Cooled Fast Breeder Reactor-Core Flow Test Loop. Thermal, mechanical, and electrical design considerations are used to derive an analytical expression that precisely describes ribbon contour in terms of the major fabrication parameters. These parameters are used to generate numerical control tapes that control ribbon cutting and winding machines. Infrared scanning techniques are developed to determine the optimum transient thermal profile of the coils and relate this profile to that generated by the coils in completed fuel pin simulators

  4. Sub-channel analysis of Pb-Bi cooled fast breeder reactor PEACER fuel assembly using MATRA

    International Nuclear Information System (INIS)

    Full text of publication follows: The nuclear power is the one of the realistic means that can solve the shortage of usable energies due to depletion of fossile fuel and due to the environmental contamination. However, since the nuclear fission yields a kind of fission fragment as a by-product, a radiological hazard of spent fuel is now a major problem. To overcome this difficulty, a number of studies are being performed and planned. One of the key solutions to this problem is to eliminate spent fuel by nuclear transmutation. According to this research, the most significant long-lived isotopes in spent fuels of the current power reactors can be transmuted into short-lived ones by using fast neutron spectrum with localized thermal traps. For this reason, liquid metal cooled fast breeder reactor is widely chosen as the answer to solve the problem. In Korea, PEACER(Proliferation-resistant, Environment-friendly, Accident-tolerant, Continual and Economical Reactor) is under study to work out this issue. PEACER core is designed to produce about 1560 MW of the thermal output with electric output up to 550 MW which efficiency is about 0.35. PEACER uses control rode made of B4C to perform reactivity control and Pb-Bi liquid metal is adopted as a coolant for primary system. In nuclear power plants it is important to keep the temperature of the reactor core structures under certain criteria in order to prevent damage of fuel materials which can advance to severe situations such as radiation leakage, and even meltdown of the fuel. This study was intended to see the liquid metal coolant behavior along the PEACER fuel channels and to find out whether the given heat flux profiles and geometrical arrangement of the fuel rods yields reasonable fluid dynamic distribution under nominal operation by using subchannel approach. The subchannel analysis of the fuel assembly under nominal operation condition was performed using MATRA (Multi-channel Analyzer for Transient and steady

  5. Fission-suppressed hybrid reactor: the fusion breeder

    International Nuclear Information System (INIS)

    Results of a conceptual design study of a 233U-producing fusion breeder are presented. The majority of the study was devoted to conceptual design and evaluation of a fission-suppressed blanket and to fuel cycle issues such as fuel reprocessing, fuel handling, and fuel management. Studies in the areas of fusion engineering, reactor safety, and economics were also performed

  6. Fission-suppressed hybrid reactor: the fusion breeder

    Energy Technology Data Exchange (ETDEWEB)

    Moir, R.W.; Lee, J.D.; Coops, M.S.

    1982-12-01

    Results of a conceptual design study of a /sup 233/U-producing fusion breeder are presented. The majority of the study was devoted to conceptual design and evaluation of a fission-suppressed blanket and to fuel cycle issues such as fuel reprocessing, fuel handling, and fuel management. Studies in the areas of fusion engineering, reactor safety, and economics were also performed.

  7. Conceptual design of a uranyl nitrate fueled reactor for the destructive testing of liquid metal fast breeder reactor fuel subassemblies

    International Nuclear Information System (INIS)

    A preliminary design of a uranyl nitrate test reactor is developed, with emphasis placed on the core neutronics and cross section development. ENDF/B-IV cross section data and the AMPX system were used to develop a 25 group neutron cross section library. A series of one-dimensional transport calculations were made in order to arrive at a reference design. Power densities of 16.5 Kw/1 appear to be attainable in the 217 pin FFTF test subassembly, with a peak neutron flux in the test zone of 2.4 x 1014 n/cm2-sec. Other engineering features pertinent to the overall system design are discussed, including: (1) corrosion, (2) treatment of radiolytic gas, (3) heat removal, and (4) reactor control

  8. 04 - Sodium cooled fast breeder fourth-generation reactors - The experimental reactor ALLEGRO, the other ways for fast breeder fourth-generation reactors

    International Nuclear Information System (INIS)

    The authors first present the technology of gas-cooled fast breeder reactors (basic principles, specific innovations, feasibility studies, fuel element, safety) and notably the ALLEGRO project (design options and expected performances, preliminary safety demonstration). Then, they present the lead-cooled fast-breeder reactor technology: interests and obstacles, return on experience, the issue of lead density, neutron assessment, transmutation potential, dosimetry, safety chemical properties and compatibility with the fuel, water, air and steels. The next part addresses the technology of molten-salt fast-breeder reactors: choice of the liquid fuel and geometry, reactor concept (difficulties, lack of past R and D), demonstration and demonstrators, international context

  9. The fast breeder reactor. v. 1

    International Nuclear Information System (INIS)

    The Energy Committee's report was prepared after hearing evidence (the minutes of which are published in Volume II) from the Central Electricity Generating Board, the United Kingdom Atomic Energy Authority and the Department of Energy. Memoranda received from other interested bodies or individuals were also considered and members of the Committee visited fast breeder projects in France, West Germany and Japan. As well as the development of the fast reactors, the economics and timescale were reviewed. The particular case of the fast breeder reactor and proposed fuel reprocessing plant at Dounreay was considered. The main conclusion is that major expenditure on fast reactor programmes can only be justified if there is a potential economic case, i.e. if the fuel cycle costs are lower than for PWRs. This would only be the case if uranium costs increased greatly. It is not considered worthwhile to participate in the European Fast Reactor although this should be reviewed in 1993 and 1997. The Committee agree with the Government's decision to cease funding the PFR in 1994 and endorses the need to regenerate the local economy which will be affected by this decision. (UK)

  10. The United States of America fast breeder reactor program

    International Nuclear Information System (INIS)

    The reasons for the development of the fast breeder reactor in the United States are outlined, and the LMFBR program is discussed in detail, under the following headings: program objectives, reactor physics, fuel and materials development, fuel recycle, safety, components, plant experience program (Near Commercial Breeder Reactor). The special facilities to be used at each stage of the program are described. It is planned that the Near Commercial Breeder Reactor will be complete in 1986, and commercial plants should follow in rapid succession. An alternate fast reactor concept (Gas Cooled Fast Reactor) is outlined. The Environmental Impact Statement for the proposed program is summarized, and the cost benefit analysis supplied as part of the Environment Statement is also summarized. (U.K.)

  11. Study on laser welding of fuel clad tubes and end plugs made of modified 9Cr-1Mo steel for metallic fuel of Fast Breeder Reactors

    Science.gov (United States)

    Harinath, Y. V.; Gopal, K. A.; Murugan, S.; Albert, S. K.

    2013-04-01

    A procedure for Pulsed Laser Beam Welding (PLBW) has been developed for fabrication of fuel pins made of modified 9Cr-1Mo steel for metallic fuel proposed to be used in future in India's Fast Breeder Reactor (FBR) programme. Initial welding trials of the samples were carried out with different average power using Nd-YAG based PLBW process. After analyzing the welds, average power for the weld was optimized for the required depth of penetration and weld quality. Subsequently, keeping the average power constant, the effect of various other welding parameters like laser peak power, pulse frequency, pulse duration and energy per pulse on weld joint integrity were studied and a procedure that would ensure welds of acceptable quality with required depth of penetration, minimum size of fusion zone and Heat Affected Zone (HAZ) were finalized. This procedure is also found to reduce the volume fraction delta-ferrite in the fusion zone.

  12. The passive nondestructive assay of the plutonium content of spent-fuel assemblies from the BN-350 fast-breeder reactor in the city of Aqtau, Kazakhstan

    International Nuclear Information System (INIS)

    The International Atomic Energy Agency is presently interested in developing equipment and techniques to measure the plutonium content of breeder reactor spent-fuel assemblies located in storage ponds before they are relocated to more secure facilities. We present the first quantitative nondestructive assay of the plutonium content of fast-breeder reactor spent-fuel assemblies while still underwater in their facility storage pond. We have calibrated and installed an underwater neutron coincidence counter (Spent Fuel Coincidence Counter (SFCC)) in the BN-350 reactor spent-fuel pond in Aqtau, Kazakhstan. A procedure has been developed to convert singles and doubles (coincidence) neutron rates observed by the SFCC into the total plutonium content of a given BN-350 spent-fuel assembly. The plutonium content has been successfully determined for spent-fuel assemblies with a contact radiation level as high as ∼105 Rads/h. Using limited facility information and multiple measurements along the length of spent-fuel assemblies, the combined measurement and facility declaration error is ∼8%. A simplified one-point measurement procedure leads to a combined measurement and facility declaration error of ∼13%

  13. Safeguards in Prototype Fast Breeder Reactor Monju

    International Nuclear Information System (INIS)

    The assemblies loaded in the core and stored in the ex-vessel storage tank (EVST) are in liquid sodium in the Japanese prototype fast breeder reactor (FBR) Monju. Since it is difficult to apply a direct verification procedure for the fuel assemblies in these areas, a dual containment and surveillance system consisting of two monitoring devices such as surveillance camera and radiation monitor that are functionally independent has been applied. In addition, the Monju Remote Monitoring System was developed to strengthen the continuous surveillance and to reduce the load of the inspection activities. Furthermore, the ex-vessel transfer machine radiation monitor (EVRM) and the exit gate monitor (EXGM) were upgraded to strengthen the monitoring of spent blanket fuel assemblies and to improve the reliability of distinguishing between fuel assemblies and non-fuel items. As the result, the integrated safeguards was introduced in November 2009, and the effective safeguards activities have been implemented in Monju. (author)

  14. Concept and development status of fast breeder reactor fuels in the FaCT project

    International Nuclear Information System (INIS)

    The fuel development and the conceptual design study have been progressed in the first phase of the FaCT project in Japan. Significant outcomes of key technologies related to fuel design, fuel properties, core materials, fuel fabrication have been provided. The prospects of these technologies have been identified. After the Fukushima accident, the research and development for reducing the amount and toxic level of radioactive wastes will be promoted more than before. These outcomes will be reflected on the future development

  15. Exploring new coolants for nuclear breeder reactors

    International Nuclear Information System (INIS)

    Breeder reactors are considered a unique tool for fully exploiting natural nuclear resources. In current Light Water Reactors (LWR), only 0.5% of the primary energy contained in the nuclei removed from a mine is converted into useful heat. The rest remains in the depleted uranium or spent fuel. The need to improve resource-efficiency has stimulated interest in Fast-Reactor-based fuel cycles, which can exploit a much higher fraction of the energy content of mined uranium by burning U-238, mainly after conversion into Pu-239. Thorium fuel cycles also offer several potential advantages over a uranium fuel cycle. The coolant initially selected for most of the FBR programs launched in the 1960s was sodium, which is still considered the best candidate for these reactors. However, Na-cooled FBRs have a positive void reactivity coefficient. Among other factors, this fundamental drawback has resulted in the canceled deployment of these reactors. Therefore, it seems reasonable to explore new options for breeder coolants. In this paper, a proposal is presented for a new molten salt (F2Be) coolant that could overcome the safety issues related to the positive void reactivity coefficient of molten metal coolants. Although it is a very innovative proposal that would require an extensive R and D program, this paper presents the very appealing properties of this salt when using a specific type of fuel that is similar to that of pebble bed reactors. The F2Be concept was studied over a typical MOX composition and extended to a thorium-based cycle. The general analysis took into account the requirements for criticality (opening the option of hybrid subcritical systems); the requirements for breeding; and the safety requirement of having a negative coolant void reactivity coefficient. A design window was found in the definition of a F2Be cooled reactor where the safety requirement was met, unlike for molten metal-cooled reactors, which always have positive void reactivity coefficients

  16. Materials accounting in a fast-breeder-reactor fuels-reprocessing facility: optimal allocation of measurement uncertainties

    International Nuclear Information System (INIS)

    This report describes the conceptual design of a materials accounting system for the feed preparation and chemical separations processes of a fast breeder reactor spent-fuel reprocessing facility. For the proposed accounting system, optimization techniques are used to calculate instrument measurement uncertainties that meet four different accounting performance goals while minimizing the total development cost of instrument systems. We identify instruments that require development to meet performance goals and measurement uncertainty components that dominate the materials balance variance. Materials accounting in the feed preparation process is complicated by large in-process inventories and spent-fuel assembly inputs that are difficult to measure. To meet 8 kg of plutonium abrupt and 40 kg of plutonium protracted loss-detection goals, materials accounting in the chemical separations process requires: process tank volume and concentration measurements having a precision less than or equal to 1%; accountability and plutonium sample tank volume measurements having a precision less than or equal to 0.3%, a shortterm correlated error less than or equal to 0.04%, and a long-term correlated error less than or equal to 0.04%; and accountability and plutonium sample tank concentration measurements having a precision less than or equal to 0.4%, a short-term correlated error less than or equal to 0.1%, and a long-term correlated error less than or equal to 0.05%. The effects of process design on materials accounting are identified. Major areas of concern include the voloxidizer, the continuous dissolver, and the accountability tank

  17. Status of fast breeder reactor development in the United States

    International Nuclear Information System (INIS)

    This document was prepared by the Office of the Program Director for Nuclear Energy, U.S. Department of Energy (USDOE). It sets forth the status and current activities for the development of fast breeder technology in the United States. In April 1977 the United States announced a change in its nuclear energy policy. Concern about the potential for the proliferation of nuclear weapons capability emerged as a major issue in considering whether to proceed with the development, demonstration and eventual deployment of breeder reactor energy systems. Plutonium recycle and the commercialization of the fast breeder were deferred indefinitely. This led to a reorientation of the nuclear fuel cycle program which was previously directed toward the commercialization of fuel reprocessing and plutonium recycle to the investigation of a full range of alternative fuel cycle technologies. Two major system evaluation programs, the Nonproliferation Alternative Systems Assessment Program (NASAP), which is domestic, and the International Nuclear Fuel Cycle Evaluation (INFCE), which is international, are assessing the nonproliferation advantages and other characteristics of advanced reactor concepts and fuel cycles. These evaluations will allow a decision in 1981 on the future direction of the breeder program. In the interim, the technologies of two fast breeder reactor concepts are being developed: the Liquid Metal Fast Breeder Reactor (LMFBR) and the Gas Cooled Fast Reactor (CFR). The principal goals of the fast breeder program are: LMFBR - through a strong R and D program, consistent with US nonproliferation objectives and anticipated national electric energy requirements, maintain the capability to commit to a breeder option; investigate alternative fuels and fuel cycles that might offer nonproliferation advantages; GCFR - provide a viable alternative to the LMFBR that will be consistent with the developing U.S. nonproliferation policy; provide GCFR technology and other needed

  18. The nuclear question at the start of the '80s: the breeder reactor

    International Nuclear Information System (INIS)

    The four newspaper articles and the letter cover the following matters: general introduction about breeder reactors and the situation in Swedish politics; visit to Dounreay to discuss breeder reactors (breeding, safety, plutonium production, radiation protection); PuO2-UO2 mixed fuel; description of breeder reactors; efficiency in use of U-235; DFR and PFR; breeder reactors in Swedish politics (arguments for and against nuclear power in general, breeder reactors in particular); discussion of the future of nuclear power in Sweden. (U.K.)

  19. Nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    A fuel assembly construction for liquid metal cooled fast breeder reactors is described in which the sub-assemblies carry a smaller proportion of parasitic material than do conventional sub-assemblies. (U.K.)

  20. Operating experience of Fast Breeder Test Reactor

    International Nuclear Information System (INIS)

    Fast Breeder Test Reactor (FBTR) is a 40 MWt / 13.2 MWe sodium cooled, loop type mixed carbide fuelled reactor. Its main aim is to gain experience in the design, construction and operation of fast reactors and to serve as an irradiation facility for development of fuel and structural material for future fast reactors. The reactor achieved first criticality in October 1985 with small indigenously designed and fabricated Mark I core (70% PuC-30% UC). The reactor power was subsequently raised in steps to 17.4 MWt by addition of Mark II fuel subassemblies (55% PuC-45% UC) and with the Mark I fuel operating at the designed linear heat rating of 400 W/cm. The turbo-generator was synchronized with the grid in July 1997. The achieved peak burn-up is 137 000 MWd/t so far without any fuel-clad failure. Presently the reactor is being operated at a nominal power of 15.7 MWt for irradiation of a test fuel subassembly of the Prototype Fast Breeder Reactor, which is coming up at Kalpakkam. It is also planned to irradiate test subassemblies made of metallic fuel for future fast reactor program. Being a small reactor, all feed back coefficients of reactivity including void coefficient are negative and hence the reactor is inherently safe. This was confirmed by carrying out physics tests. The capability to remove decay heat under various incidental conditions including natural convection was demonstrated by carrying out engineering tests. Thermo couples are provided for on-line monitoring of fuel SA outlet temperature by dedicated real time computer and processed to generate trip signals for the reactor in case of power excursion, increase in clad hot spot temperature and subassembly flow blockage. All pipelines and capacities in primary main circuit are provided with segmented outer envelope to minimize and contain radioactive sodium leak while ensuring forced cooling through reactor to remove decay heat in case of failure of primary boundary. In secondary circuit, provision is

  1. Investigations by model theory of fast breeder reactor fuel pins and application to special safety experiments

    International Nuclear Information System (INIS)

    This paper makes a contribution to the development of the fuel rod model theory for describing transient loads up to and including very fast accidents. In three successive sections the state of knowledge is subjected to critical discussion, improvements are presented and, finally, possibilities of application indicated by the example of experiment analysis. Within the framework of further developments a consistent compilation is given of all relevant material data for describing UC and (U, Pu) C-fuels and models are derived on the fission gas behavior and on swelling. A literature search is made on the restructuring of oxide fuels and it is shown above all with respect to transient calculations which models can be employed here. (orig./RW)

  2. Preparation of LWBR [Light Water Breeder Reactor] spent fuel for shipment to ICPP [Idaho Chemical Processing Plant] for long term storage (LWBR Development Program)

    International Nuclear Information System (INIS)

    After successfully operating for 29,047 effective full power hours, the Light Water Breeder Reactor (LWBR) core was defueled prior to total decommissioning of the Shippingport facility. All nuclear fuel and much of the reactor internal hardware was removed from the reactor vessel. Non-fuel components were prepared for shipment to disposal sites, and the fuel assemblies were partially disassembled and shipped to the Expended Core Facility (ECF) in Idaho. At ECF, the fuel modules underwent further disassembly to provide fuel rods for nondestructive testing to establish the core's breeding efficiency and to provide core components for examinations to assess their performance characteristics. This report presents a basic description of the processes and equipment used to prepare and to ship all LWBR nuclear fuel to the Idaho Chemical Processing Plant (ICPP) for long-term storage. Preparation processes included the underwater loading of LWBR fuel into storage liners, the sealing, dewatering and drying of the storage liners, and the final pressurization of the storage liners with inert neon gas. Shipping operations included the underwater installation of the fuel loaded storage liner into the Peach Bottom shipping cask, cask removal from the waterpit, cask preparations for shipping, and cask shipment by tractor trailer to the ICPP facility for long-term storage. The ICPP facility preparations for LWBR fuel storage and the ICPP process for discharge of the fuel into underground silos are presented. 10 refs., 42 figs

  3. Evaluation of molten fuel containment concepts for gas-cooled fast breeder reactors

    International Nuclear Information System (INIS)

    Four in-vessel molten fuel containment concepts for the GCFR were compared, namely, (1) a ceramic crucible, (2) a borax bath, (3) a heavy metal bath, and (4) a steel bath. The ceramic crucible is the simplest but depends on substantial upward heat removal. The borax bath and the heavy metal bath concepts offer better performance but would require design changes and an increased experimental effort. The steel bath concept is a good compromise and has potential for further improvement by combining it with the essential features of other concepts, i.e., the crucible or the heavy metal bath. It is concluded that several concepts could potentially exploit the normally provided cooled liner barrier in the PCRV cavity for post-accident fuel containment

  4. Electrolytic destruction of nitric acid in various reprocessing streams of fast breeder reactor fuels

    International Nuclear Information System (INIS)

    The salting concentration of Nitric acid to Reprocessing Plants of FBR fuels is 4 M HNO3,. Adjustment of the free acidity of the Dissolver solution from 8 to 10 M HNO3, to a concentration of 4M HNO3 is done cathodically. This would increase the throughput of the plant. The concentration of HNO3 in first cycle raffinate, HAW, that would be stored in SS tanks (interim storage before vitrification) should be less than 6 M; otherwise severe corrosion problems would occur. The HAW discharged in FBR fuel reprocessing plants would contain around 4M HNO3. This has to be reduced to around 0.4M HNO3 in order to effect efficient reduction in waste volume, by evaporation. This has been achieved in the catalytic reduction of HNO3 to oxides of nitrogen. Current efficiency in the 1 L level is around 46%. Conventionally, HNO3 in raffinates are destroyed by addition of formaldehyde. The disadvantages of this method are highlighted in this paper. Addition of 0.01 M Cu2+ is found to completely avoid critical concentration of HNO3 below which nitrate ions will not be destroyed; instead H2 gas would be evolved. The results of these experiments in a simulated HAW are also included in this paper. Operation and the results of electrolytic destruction of HNO3 in simulated raffinate in a scaled-up SS equipment of 3.5 L capacity are also described in this payer. (author)

  5. AB INITIO STUDY OF ADVANCED METALLIC NUCLEAR FUELS FOR FAST BREEDER REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Landa, A; Soderlind, P; Grabowski, B; Turchi, P A; Ruban, A V; Vitos, L

    2012-04-23

    Density-functional formalism is applied to study the ground state properties of {gamma}-U-Zr and {gamma}-U-Mo solid solutions. Calculated heats of formation are compared with CALPHAD assessments. We discuss how the heat of formation in both alloys correlates with the charge transfer between the alloy components. The decomposition curves for {gamma}-based U-Zr and U-Mo solid solutions are derived from Ising-type Monte Carlo simulations. We explore the idea of stabilization of the {delta}-UZr{sub 2} compound against the {alpha}-Zr (hcp) structure due to increase of Zr d-band occupancy by the addition of U to Zr. We discuss how the specific behavior of the electronic density of states in the vicinity of the Fermi level promotes the stabilization of the U{sub 2}Mo compound. The mechanism of possible Am redistribution in the U-Zr and U-Mo fuels is also discussed.

  6. Coincidence measurements of FFTF breeder fuel subassemblies

    International Nuclear Information System (INIS)

    A prototype coincidence counter developed to assay fast breeder reactor fuel was used to measure four fast-flux test facility subassemblies at the Hanford Engineering Development Laboratory in Richland, Washington. Plutonium contents in the four subassemblies ranged between 7.4 and 9.7 kg with corresponding 240Pu-effective contents between 0.9 and 1.2 kg. Large count rates were observed from the measurements, and plots of the data showed significant multiplication in the fuel. The measured data were corrected for deadtime and multiplication effects using established formulas. These corrections require accurate knowledge of the plutonium isotopics and 241Am content in the fuel. Multiplication-corrected coincidence count rates agreed with the expected count rates based on spontaneous fission-neutron emission rates. These measurements indicate that breeder fuel subassemblies with 240Pu-effective contents up to 1.2 kg can be nondestructively assayed using the shift-register electronics with the prototype counters. Measurements using the standard Los Alamos National Laboratory shift-register coincidence electronics unit can produce an assay value accurate to +-1% in 1000 s. The uncertainty results from counting statistics and deadtime-correction errors. 3 references, 8 figures, 8 tables

  7. Breeder Spent Fuel Handling Program multipurpose cask design basis document

    International Nuclear Information System (INIS)

    The Breeder Spent Fuel Handling (BSFH) Program multipurpose cask Design Basis Document defines the performance requirements essential to the development of a legal weight truck cask to transport FFTF spent fuel from reactor to a reprocessing facility and the resultant High Level Waste (HLW) to a repository. 1 ref

  8. Unusual occurrences in fast breeder test reactor

    International Nuclear Information System (INIS)

    Fast Breeder Test Reactor (FBTR) is a 40 MWt/13.2 MWe sodium cooled mixed carbide fuelled reactor. Its main aim is to generate experience in the design, construction and operation of fast reactors including sodium systems and to serve as an irradiation facility for the development of fuel and structural materials for future fast reactors. It achieved first criticality in Oct 85 with Mark I core (70% PuC - 30% UC). Steam generator was put in service in Jan 93 and power was raised to 10.5 MWt in Dec 93. Turbine generator was synchronised to the grid in Jul 97. The indigenously developed mixed carbide fuel has achieved a burnup of 44,000 MW-d/t max at a linear heat rating of 320 W/cm max without any fuel clad failure. The commissioning and operation of sodium systems and components have been smooth and performance of major components, viz., sodium pumps, intermediate heat exchangers and once through sodium heated steam generators (SG) have been excellent. There have been three minor incidents of Na/NaK leaks during the past 14 years, which are described in the paper. There have been no incident of a tube leak in SG. However, three incidents of water leaks from water / steam headers have been detailed. The plant has encountered some unusual occurrences, which were critically analysed and remedial measures, in terms of system and procedural modifications, incorporated to prevent recurrence. This paper describes unusual occurrences of fuel handling incident of May 1987, main boiler feed pump seizure in Apr 1992, reactivity transients in Nov 1994 and Apr 1995, and malfunctioning of the core cover plate mechanism in Jul 1995. These incidents have resulted in long plant shutdowns. During the course of investigation, various theoretical and experimental studies were carried out for better understanding of the phenomena and several inspection techniques and tools were developed resulting in enriching the technology of sodium cooled reactors. FBTR has 36 neutronic and process

  9. Integral measurement of fission products capture in fast breeder reactors

    International Nuclear Information System (INIS)

    For the SUPERPHENIX reactor project, it was necessary to know fission products capture with about 10% accuracy in the fast breeder reactor spectra. In this purpose, integral measurements have been carried out on the main separated products by different experimental technics (oscillation, activation and irradiation methods), but particularly on irradiated fuel pins from RAPSODIE and PHENIX reactors in order to directly obtain total effect of fission products. Same tendencies have been observed for both enriched uranium fuel and LMFBR characteristic plutonium fuel. All experimental results have been introduced in CARNAVAL cross section set

  10. Status and prospects of thermal breeders and their effect on fuel utilization

    International Nuclear Information System (INIS)

    The report evaluates the extent to which thermal breeders and near-breeders might complement fast breeders or serve as an alternative in solving the long-term nuclear fuel supply problem. It considers in a general way issues such as proliferation, safety, environmental impacts, economics, power plant availability and fuel cycle versatility in order to determine whether thermal breeder reactors offer advantages or disadvantages with respect to such issues

  11. Improved analysis on multiple recycling of fuel in prototype fast breeder reactor in a closed fuel cycle

    Indian Academy of Sciences (India)

    G Pandikumar; V Gopalakrishnan; P Mohanakrishnan

    2011-08-01

    An FBR closed fuel cycle involves recycling of the discharge fuel, after reprocessing and refabrication, to utilize the unburnt fuel remains and the freshly bred fissile material. Our previous study in this regard for the PFBR indicated a comfortable feasibility of multiple recycling with selfsufficiency. In the present work, more refined estimations are done using the most recent nuclear data, viz. ENDF/B-VII.0, and with the most recent specification of the fuel composition. Among others, this paper brings out the importance of taking into account the energy self-shielding effects in the cross-section averages used in the study. While self-shielded averages lead to realistic predictions, unshielded averages significantly overpredict breeding in the blankets and underpredict loss in the cores.

  12. The fast breeder reactor Rapsodie (1962)

    International Nuclear Information System (INIS)

    In this report, the authors describe the Rapsodie project, the French fast breeder reactor, as it stands at construction actual start-up. The paper provides informations about: the principal neutronic and thermal characteristics, the reactor and its cooling circuits, the main handling devices of radioactive or contaminated assemblies, the principles and means governing reactor operation, the purposes and locations of miscellaneous buildings. Rapsodie is expected to be critical by 1964. (authors)

  13. Experimental Breeder Reactor I Preservation Plan

    Energy Technology Data Exchange (ETDEWEB)

    Julie Braun

    2006-10-01

    Experimental Breeder Reactor I (EBR I) is a National Historic Landmark located at the Idaho National Laboratory, a Department of Energy laboratory in southeastern Idaho. The facility is significant for its association and contributions to the development of nuclear reactor testing and development. This Plan includes a structural assessment of the interior and exterior of the EBR I Reactor Building from a preservation, rather than an engineering stand point and recommendations for maintenance to ensure its continued protection.

  14. Coatings for fast breeder reactor components

    International Nuclear Information System (INIS)

    Several types of metallurgical coatings are used in the unique environments of the fast breeder reactor. Most of the coatings have been developed for tribological applications, but some also serve as corrosion barriers, diffusion barriers, or radionuclide traps. The materials that have consistently given the best performance as tribological coatings in the breeder reactor environments have been coatings based on chromium carbide, nickel aluminide, or Tribaloy 700 (a nickel-base hard-facing alloy). Other coatings that have been qualified for limited applications include chromium plating for low temperature galling protection and nickel plating for radionuclide trapping

  15. FUJI, an initial sintering comparison test for pelletized-, sphere-pac- and vipac- fast breeder reactor mixed oxide fuel

    International Nuclear Information System (INIS)

    Options for fuel cycle technology improvement have strongly regained attention lately with the revival of nuclear energy production interests and plants for next generation nuclear systems. Various fuel forms, geometries and production paths are being looked at. Within the FUJI collaboration program among, Japan Nuclear Cycle Development Institute (JNC, Japan), Paul Scherrer Institute (PSI, Switzerland) and Nuclear Research and Consultancy Group (NRG, the Netherlands) the production paths of plutonium and neptunium mixed oxide- (sphere-pac- and vipac-) particle fuels (20wt% Pu and 5wt% Np) are tested as well as initial sintering and power-to-melt environment under simulated FBR conditions. The various fuel forms were produced at PSI under the support of JNC, the irradiations were accomplished at High Flux Reactor (HFR) in Petten, the post irradiation examinations are being achieved mainly at NRG and the fuel modelling being performed at JNC and PSI. The present paper reviews the project planning, fuel behaviour- pre-calculations and the fuel- and fuel segment- production, while a second paper at this conference summarizes the reactor irradiations and the status of the available post irradiation examination results. (author)

  16. Status of the DEBENE fast breeder reactor development, March 1979

    International Nuclear Information System (INIS)

    Status report of the Fast-breeder reactor development in Germany covers the following: description of the political situation in Federal republic of germany during 1978; international cooperation in the field of fast reactor technology development; operation description of the KNK-II fast core experimental power plant; status of construction of the SNR-300; results of the research and development programs concerned with fuel element, cladding, absorber rods and core structural materials development; sodium effects; neutron irradiation effects on SS properties; reactor physics related to experiments in fast critical assemblies; fast reactor safety issues; core disruption accidents; sodium boiling experiments, measuring methods developed; component tests

  17. Analysis of a sustainable gas cooled fast breeder reactor concept

    International Nuclear Information System (INIS)

    Highlights: • A Thorium-GFBR breeder for actinide recycling ability, and thorium fuel feasibility. • A mixture of 232Th and 233U is used as fuel and LWR used fuel is used. • Detailed neutronics, fuel cycle, and thermal-hydraulics analysis has been presented. • Run this TGFBR for 20 years with breeding of 239Pu and 233U. • Neutronics analysis using MCNP and Brayton cycle for energy conversion are used. - Abstract: Analysis of a thorium fuelled gas cooled fast breeder reactor (TGFBR) concept has been done to demonstrate the self-sustainability, breeding capability, actinide recycling ability, and thorium fuel feasibility. Simultaneous use of 232Th and used fuel from light water reactor in the core has been considered. Results obtained confirm the core neutron spectrum dominates in an intermediate energy range (peak at 100 keV) similar to that seen in a fast breeder reactor. The conceptual design achieves a breeding ratio of 1.034 and an average fuel burnup of 74.5 (GWd)/(MTHM) . TGFBR concept is to address the eventual shortage of 235U and nuclear waste management issues. A mixture of thorium and uranium (232Th + 233U) is used as fuel and light water reactor used fuel is utilized as blanket, for the breeding of 239Pu. Initial feed of 233U has to be obtained from thorium based reactors; even though there are no thorium breeders to breed 233U a theoretical evaluation has been used to derive the data for the source of 233U. Reactor calculations have been performed with Monte Carlo radiation transport code, MCNP/MCNPX. It is determined that this reactor has to be fuelled once every 5 years assuming the design thermal power output as 445 MW. Detailed analysis of control rod worth has been performed and different reactivity coefficients have been evaluated as part of the safety analysis. The TGFBR concept demonstrates the sustainability of thorium, viability of 233U as an alternate to 235U and an alternate use for light water reactor used fuel as a blanket for

  18. Blanket management method for liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    A method for reducing thermal striping in liquid metal fast breeder reactors by reducing temperature gradients between adjacent fuel and blanket assemblies by shuffling blanket assemblies at each refueling outage so as to progressively shuffle the blanket assemblies to the core periphery through multiple moves and to generally locate fresh blanket assemblies adjacent to exposed fuel assemblies and exposed blanket assemblies adjacent to fresh fuel. Additionally, assembly orificing is altered to provide less flow to blanket assemblies needing less flow due to an otherwise decreased temperature gradient and providing additional flow to fuel assemblies which need more flow to sufficiently reduce temperature gradients to prevent thermal striping. (author)

  19. CAMDYN: a new model to describe the axial motion of molten fuel inside the pin of a fast breeder reactor during accident conditions

    International Nuclear Information System (INIS)

    The new in-pin fuel motion model CAMDYN (Cavity Material Dynamics) describes the axial motion of both partially and fully molten fuel inside the pin of a fast breeder reactor during accident conditions. The motion of the two types of molten fuel and the imbedded fission gas bubbles is treated both before and after cladding failure. The basic modelling approach consists of the treatment of two one-dimensional flows which are coupled by interaction terms. Each of these flows is treated compressively and with axially variable flow cross sections. The mass and energy equations of both fields are solved explicitly using upwind differencing on a fixed Eulerian grid. The two momentum equations are solved simultaneously, using the convective momentum fluxes of the previous timestep. Both partially and fully molten fuel can move axially into a central hole extending to the plenum in the case of certain hollow pellet designs. The fuel temperature calculation includes the determination of a radial temperature profile. A simple conduction freezing model is included. After cladding failure, ejection into the coolant channel is modeled

  20. Liquid metal fast breeder reactor: an environmental and economic critique

    International Nuclear Information System (INIS)

    Economic and environmental arguments made by the AEC and others for the liquid metal fast breeder reactor (LMFBR) as a central component of the U. S. electrical energy system are discussed. The LMFBR appears to have no environmental advantage over the currently operating light water reactor and especially not over the high temperature gas reactor. The principle environmental argument for the rapid introduction of LMFBRs is that they will provide a virtually inexhaustible fuel source, and reduce the demand for strip-mining the limited reserves of high grade U ore. A 20-yr delay in the construction of LMFBRs would result in an increase of only 50 mi2 of strip mining over the next 50 yr, and the cost of reclamation of this land would be about 0.1 mill/kw-hr. Uranium from which fuel has been extracted for use by nonbreeder reactors can still be used by breeders, thus breeders could still be introduced in the future, if fusion is not developed in time, and extract the same overall energy from a given supply of U as if they had been introduced earlier. Economic arguments in favor of the LMFBR are based on models highly sensitive to changes on some of the most critical input variables: nuclear power plant capital costs, fuel cycle costs, performance characteristics of LMFBR designs, electrical energy demand, and U ore costs. There is no basis for concluding that the LMFBR will be economical in the 1980s or early 1990s. (Pollut. Abstr.)

  1. The Shippingport Pressurized Water Reactor and Light Water Breeder Reactor

    International Nuclear Information System (INIS)

    This report discusses the Shippingport Atomic Power Station, located in Shippingport, Pennsylvania, which was the first large-scale nuclear power plant in the United States and the first plant of such size in the world operated solely to produce electric power. A program was started in 1953 at the Bettis Laboratory to confirm the practical application of nuclear power for large-scale electric power generation. It led to the development of zirconium alloy (Zircaloy) clad fuel element containing bulk actinide oxide ceramics (UO2, ThO2, ThO2 -- UO2, ZrO2 -- UO2) as nuclear reactor fuels. The program provided much of the technology being used for design and operation of the commercial, central-station nuclear power plants now in use. The Shippingport Pressurized Water Reactor (PWR) began initial power operation on December 18, 1957, and was a reliable electric power producer until February 1974. In 1965, subsequent to the successful operation of the Shippingport PWR (UO2, ZrO2 -- UO2 fuels), the Bettis Laboratory undertook a research and development program to design and build a Light Water Breeder Reactor (LWBR) core for operation in the Shippingport Station. Thorium was the fertile fuel in the LWBR core and was the base oxide for ThO2 and ThO2 -- UO2 fuel pellets. The LWBR core was installed in the pressure vessel of the original Shippingport PWR as its last core before decommissioning. The LWBR core started operation in the Shippingport Station in the autumn of 1977 and finished routine power operation on October 1, 1982. Successful LWBR power operation to over 160% of design lifetime demonstrated the performance capability of the core for both base-load and swing-load operation. Postirradiation examinations confirmed breeding and successful performance of the fuel system

  2. Instrumentation and control improvements at Experimental Breeder Reactor II

    Energy Technology Data Exchange (ETDEWEB)

    Christensen, L.J.; Planchon, H.P.

    1993-01-01

    The purpose of this paper is to describe instrumentation and control (I C) system improvements at Experimental Breeder Reactor 11 (EBR-11). The improvements are focused on three objectives; to keep the reactor and balance of plant (BOP) I C systems at a high level of reliability, to provide diagnostic systems that can provide accurate information needed for analysis of fuel performance, and to provide systems that will be prototypic of I C systems of the next generation of liquid metal reactor (LMR) plants.

  3. Instrumentation and control improvements at Experimental Breeder Reactor II

    Energy Technology Data Exchange (ETDEWEB)

    Christensen, L.J.; Planchon, H.P.

    1993-03-01

    The purpose of this paper is to describe instrumentation and control (I&C) system improvements at Experimental Breeder Reactor 11 (EBR-11). The improvements are focused on three objectives; to keep the reactor and balance of plant (BOP) I&C systems at a high level of reliability, to provide diagnostic systems that can provide accurate information needed for analysis of fuel performance, and to provide systems that will be prototypic of I&C systems of the next generation of liquid metal reactor (LMR) plants.

  4. Binary breeder reactor: an option for Brazilian energy future

    International Nuclear Information System (INIS)

    To assure a continued supply of electric energy beyond a few decades from now, developmemnt of fast breeder reactors is a necessity. Binary fueled LMFBRs combine an improvement in the inherent safety of fast reactors and an efficient use of the abundant thorium. A nuclear system that starts with PWRs and gradually shifts to a FBR system or to a FBR-PWR symbiotic system appears to be the most reasonable one for Brazil. Natural uranium requirements and possible sequences of reactor introductions are discussed for some postulated Brazilian situations. A permanent system of approx. 100 GWe capacity can be established based on the estimated resource of natural uranium. (Author)

  5. Binary breeder reactor an option for Brazilian energy future

    International Nuclear Information System (INIS)

    To assure a continued supply of electric energy beyond a few decades from now, development of fast breeder reactors is a necessity. Binary fueled LMFBRs combine an improvement in the inherent safety of fast reactors and an efficient use of the abundant thorium. A nuclear system that starts with PWRs and gradually shifts to a FBR system or to a FBR-PWR symbiotic system appears to be the most resonable one for Brazil. Natural uranium requirements and possible sequences of reactor introductions are discussed for some postulated Brazilian situations. A permanent system of approximatelly 100 GWe capacity can be established based on the estimated resource of natural uranium. (Author)

  6. A fast breeder reactor development scheme for Brazil

    International Nuclear Information System (INIS)

    Fast breeder reactors will be necessary in the next century in order to meet increasing demands for electricity resulting from industrialization and general improvement of standards of living. A scheme for the development of liquid metal fast breeder reactors in Brazil is proposed. Emphasis are placed on reactor safety in order to promote public acceptance, on utilization of thorium that is abundant in the country, and on consistency and smoothness of the development. The initial step is the construction and operation of a 5 MW experimental fast reactor in order to acquire basic experiences and technologies. The second step is the construction of a series of small power plants which should assure a ssound technological development. The reactor is designed with particular emphasis on safety and ease of operation. Demonstration of safety and reliability with small units would enhance public acceptance. In the final phase, when fast breeder reactors are to play a central role in electricity generation, large power plants that utilize both uranium and thorium fuel cycles will be built to establish a practically permanent power system. (Author)

  7. The breeder reactor in electricity supply

    International Nuclear Information System (INIS)

    Forecasts are made of Britain's energy prospects in the year 2000. It is concluded that fossil fuels and renewable energy sources will leave an energy gap and that dependence on nuclear power will be substantial. There will, however have been a rapid depletion of readily available uranium ore reserves and a growing availability of plutonium from thermal reactors. Britain's resources of plutonium and depleted uranium which the fast breeder reactor can use - will equal many thousand million tonnes of coal. Our nuclear programme should therefore include one or two FBRs. Resources should be pooled internationally and plants built to prove alternative options and consolidate an already highly developed technology. In Britain our earliest nuclear (Magnox) stations have served as well. In Scotland, where next year an estimated 30% of electricity output will be nuclear, Hunterston 'B' AGR has had a splendid first year of operation, and pumped storage capacity in Scotland has extended the benefits of low-cost generation. The FBR has many very satisfactory engineering features and is eminently controllable and well behaved. It is compact, with relatively low cooling-water requirements and it could be built, one hopes, close to our load centres. There can be confidence that it will be proved safe. An order for an FBR, on the earliest timescale that fits in with evidence of successful operation of the Dounreay PFR and an agreed international programme, would serve the national interest and ensure the survival of our plant manufacturers, so badly hit by the effects of stagnation of the U.K. economy. (author)

  8. Status of fast breeder reactor development in the United States

    International Nuclear Information System (INIS)

    The energy policy of the United States is aimed at shifting as rapidly as practicable from an oil dependent economy to one that relies heavily on other fuels and energy sources. Nuclear power Is now and is expected to continue to be an important factor in achieving this goal. If nuclear power is to contribute to a solution of future energy needs, demonstration of the breeder reactor as a viable source of essentially inexhaustible energy supply is essential. The US DOE program for development of the fast breeder reactor has witnessed some notable events in the past year. Foremost among these Is the successful operational testing of the Fast Flux Test Facility (FFTF), located at.the Hanford Engineering Development Laboratory. The reactor reached full design power of 400 MW(t) on December 21, 1980, and has performed remarkably close to design specifications. Design of the Clinch River Breeder Reactor Plant (CRBRP), a 375 MW(e) LMFBR, is now over 80 percent complete. About $530 million in components have been ordered; component deliveries total approximately $124 million; work-in-process totals another $204 million. Construction of the plant, however, has been suspended since 1977. With the concurrence of the U.S. Congress and approvals from the appropriate authorities work on the safety review and site clearing for construction can resume. The Conceptual Design Study for a large, 1000 MW(e) LMFBR Large Developmental Plant was recently completed on a schedule commensurate with submission of a full report to the Congress at the end of March, 1981. This report is the culmination of a study which began in October, 1978 and involved contributions from U.S. reactor manufacturers and US DOE laboratories. The US DOE is carrying forward a comprehensive technology development program. This effort provides direct support to the FFTF and CRBRP projects and to the LDP. It also supports technology development which is generic to the overall LMFBR program. Funding for breeder

  9. Installation of the Light-Water Breeder Reactor at the Shippingport Atomic Power Station (LWBR Development Program)

    International Nuclear Information System (INIS)

    This report summarizes the refueling operations performed to install a Light Water Breeder Reactor (LWBR) core into the existing pressurized water reactor vessel at the Shippingport Atomic Power Station. Detailed descriptions of the major installation operations (e.g., primary system preconditioning, fuel installation, pressure boundary seal welding) are included as appendices to this report; these operations are of technical interest to any reactor servicing operation, whether the reactor is a breeder or a conventional light water non-breeder core

  10. Installation of the Light-Water Breeder Reactor at the Shippingport Atomic Power Station (LWBR Development Program)

    Energy Technology Data Exchange (ETDEWEB)

    Massimino, R.J.; Williams, D.A.

    1983-05-01

    This report summarizes the refueling operations performed to install a Light Water Breeder Reactor (LWBR) core into the existing pressurized water reactor vessel at the Shippingport Atomic Power Station. Detailed descriptions of the major installation operations (e.g., primary system preconditioning, fuel installation, pressure boundary seal welding) are included as appendices to this report; these operations are of technical interest to any reactor servicing operation, whether the reactor is a breeder or a conventional light water non-breeder core.

  11. Liquid Metal Fast Breeder Reactor program. Volume III. Environmental statement

    International Nuclear Information System (INIS)

    The various alternative technologies, nuclear as well as nonnuclear, that might be utilized in conjunction with or instead of the LMFBR to satisfy the Nation's future electric power requirements are examined. The options considered include the further implementation of various types of nuclear power reactors such as the already existing light water reactor and high temperature gas-cooled reactor, as well as the development of alternative breeder reactors such as the gas-cooled fast reactor, light water breeder reactor and molten salt breeder reactor. The development of another potential nuclear energy system, controlled thermonuclear fusion, is also addressed. The possibilities of increased emphasis on the use of conventional fossil fuels, namely coal, oil and natural gas, and the development of unconventional fossil fuels such as oil shale and domestic tar sands are discussed, followed by consideration of the further development of additional nonnuclear energy sources such as hydroelectric power systems, geothermal energy, solar energy, and other potential sources of power. Each option is examined as to the extent of its energy resource base, the research and development program that would be required (if any) to bring the option into commercial use, the environmental implications of its utilization and the costs and benefits associated with its use, in order to assess its capability for satisfying projected energy requirements. The use of improved energy conversion and storage devices such as gas turbines, fuel cells and magnetohydrodynamics is discussed. An examination of the various elements of a potential national effort in energy conservation to assess their capabilities for reducing projected energy demands and thereby replacing partially or entirely the need for additional power sources such as the LMFBR is presented. (U.S.)

  12. CALIPSO - a computer code for the calculation of fluiddynamics, thermohydraulics and changes of geometry in failing fuel elements of a fast breeder reactor

    International Nuclear Information System (INIS)

    The computer code CALIPSO was developed for the calculation of a hypothetical accident in an LMFBR (Liquid Metal Fast Breeder Reactor), where the failure of fuel pins is assumed. It calculates two-dimensionally the thermodynamics, fluiddynamics and changes in geometry of a single fuel pin and its coolant channel in a time period between failure of the pin and a state, at which the geometry is nearly destroyed. The determination of temperature profiles in the fuel pin cladding and the channel wall make it possible to take melting and freezing processes into account. Further features of CALIPSO are the variable channel cross section in order to model disturbances of the channel geometry as well as the calculation of two velocity fields including the consideration of virtual mass effects. The documented version of CALIPSO is especially suited for the calculation of the SIMBATH experiments carried out at the Kernforschungszentrum Karlsruhe, which simulate the above-mentioned accident. The report contains the complete documentation of the CALIPSO code: the modeling of the geometry, the equations used, the structure of the code and the solution procedure as well as the instructions for use with an application example. (orig.)

  13. Safety and core design of large liquid-metal cooled fast breeder reactors

    OpenAIRE

    Qvist, Staffan Alexander

    2013-01-01

    In light of the scientific evidence for changes in the climate caused by greenhouse-gas emissions from human activities, the world is in ever more desperate need of new, inexhaustible, safe and clean primary energy sources. A viable solution to this problem is the widespread adoption of nuclear breeder reactor technology. Innovative breeder reactor concepts using liquid-metal coolants such as sodium or lead will be able to utilize the waste produced by the current light water reactor fuel cyc...

  14. Optimization of U–Th fuel in heavy water moderated thermal breeder reactors using multivariate regression analysis and genetic algorithms

    International Nuclear Information System (INIS)

    Highlights: • A new method useful for the parametric analysis and optimization of reactor core designs. • This uses the strengths of genetic algorithms (GA), and regression splines. • The method is applied to the core fuel pin cell of a PHWR design. • Tools like java, R, and codes like Serpent, Matlab are used in this research. - Abstract: An analysis and optimization of a set of neutronics parameters of a thorium-fueled pressurized heavy water reactor core fuel has been performed. The analysis covers a detailed pin-cell analysis of a seed-blanket configuration, where the seed is composed of natural uranium, and the blanket is composed of thorium. Genetic algorithms (GA) is used to optimize the input parameters to meet a specific set of objectives related to: infinite multiplication factor, initial breeding ratio, and specific nuclide’s effective microscopic cross-section. The core input parameters are the pitch-to-diameter ratio, and blanket material composition. Recursive partitioning of decision trees (rpart) multivariate regression model is used to perform a predictive analysis of the samples generated from the GA module. Reactor designs are usually complex and a simulation needs a significantly large amount time to execute, hence implementation of GA or any other global optimization techniques is not feasible, therefore we present a new method of using rpart in conjunction with GA. Due to using rpart, we do not necessarily need to run the neutronics simulation for all the inputs generated from the GA module rather, run the simulations for a predefined set of inputs, build a regression fit to the input and the output parameters, and then use this fit to predict the output parameters for the inputs generated by GA. The rpart model is implemented as a library using R programming language. The results suggest that the initial breeding ratio tends to increase due to a harder neutron spectrum, however a softer neutron spectrum is desired to limit the

  15. The breeder spent fuel packaging and transportation program

    International Nuclear Information System (INIS)

    The Breeder Spent Fuel Handling and Transportation Program of the United States Department of Energy (DOE) was established in 1983 in order to develop a reliable planning base for interface development at the back end of the liquid metal fast breeder reactor (LMFBR) fuel cycle. It began by addressing the immediate interface needs between the planned Clinch River Breeder Reactor, near Oak Ridge, Tennessee, and the proposed Breeder Reprocessing Engineering Test Facility at Richland, Washington, and concluded by providing a developmental plan leading to a sodium-cooled spent breeder fuel transportation cask for a mature 20-reactor LMFBR industry in the year 2025. During the formulation of this plan, as well as during the technology development that constituted the programme, liaison between the DOE and the concerned private industry operations was maintained by frequent meetings. As a result of functional considerations, it was decided that a legal truck-weight stainless steel multi-assembly package would both be economical and would have unlimited routine possibilities and facility access. As the detailed conceptual design emerged, it included remotely workable, spring-loaded, captive bolts to reduce occupational exposure, internal integral impact limiters and a structurally promising depleted uranium gamma shield. Modular baskets of a boron-aluminium alloy, produced by Fonderies Montupet of France, would enhance criticality control and heat transfer, as well as allowing for either a spent fuel or high level waste payload. While preliminary calculations have qualified the structure and shielding, heat transfer from a six-assembly payload still poses problems. Details are discussed in the paper. (author)

  16. Fast breeder reactors: Experience and trends. V. 2

    International Nuclear Information System (INIS)

    The IAEA Symposium on ''Fast Breeder Reactors: Experience and Future Trends'' was held, at the invitation of the Government of France, in Lyons, France, on 22-26 July 1985. It was hosted by the French Commissariat a l'energie atomique and Electricite de France. The purpose of the Symposium was to review the experience gained so far in the field of LMFBRs, taking into account the constructional, operational, technological, economic and fuel cycle aspects, and to consider the developmental trends as well as the international co-operation in fast breeder reactor design and utilization. The Symposium was attended by almost 400 participants (340 participants, 35 observers and 20 journalists) from 25 countries and five international organizations. More than 80 papers were presented and discussed during six regular sessions and four poster sessions. A separate abstract was prepared for each of these papers

  17. Feasibility and deployment strategy of water cooled thorium breeder reactor

    International Nuclear Information System (INIS)

    The author have studied water cooled thorium breeder reactor based on matured pressurized water reactor (PWR) plant technology for several years. Through these studies it is concluded that reduced moderated core by arranging fuel pins in a triangular tight lattice array with heavy water coolant in the primary loop by replacing original light water is appropriate for achieving sufficient breeding performance as sustainable fission system and high enough burn-up as an economical power plant. The heavy water cooled thorium reactor is feasible to be introduced by using Pu recovered from spent fuel of LWR, keeping continuity with current LWR infrastructure. This thorium reactor can be operated as sustainable energy supplier and also MA transmuter to realize future society with less long-lived nuclear waste

  18. Thermal and neutronic calculation for fast breeder reactor FBR

    International Nuclear Information System (INIS)

    This research included studying of thermal and neutronic calculation for fast breeder nuclear reactor, to putting the optimum design for this reactor. So a Soviet type (BN-350) was chosen, which has its core composed of two enrichment zones, and with blanket that contains depleted uranium. A group of thermal calculation programs was made by using personal computer, to obtain core and blanket reactor dimensions and volume fractions of reaction input material and number and dimensions of fuel rods which were used for neutron calculations. Several core and blanket enrichments were used to study neutron flux behaviour for two reactors different conditions. First when control rods exist in the core reactor and second when the rods are out of the core. Breeding ratio was also studied for different core and blanket enrichment. 30 tabs.; 24 figs.; 34 refs.; 3 apps

  19. Technological questions of the breeder fuel cycle

    International Nuclear Information System (INIS)

    Since the contributions by the Karlsruhe Nuclear Research Center to the construction of SNR 300 have been completed to a large extent and irradiated KNK II fuel subassemblies have now become available, the possibility and necessity arise of concentrating efforts on the breeder fuel cycle. This work was started in 1980. The 17 papers presented at this seminar will provide a survey of intermediate results obtained until today. (orig./HP)

  20. Degrading the Plutonium Produced in Fast Breeder Reactor Blankets

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jor-Shan; Kuno, Yusuke [Tokyo University, 7-3-1, Hongo, Bunkyo-ku, Tokyo, 113-8656 (Japan)

    2009-06-15

    Plutonium quality, defined as the plutonium isotopic composition, is an important measure for proliferation-resistance (PR) of a nuclear energy system. The quality of the plutonium produced in the blanket assemblies of a fast breeder reactor could be as good as or better than the weapons-grade (WG). The presence of such good quality plutonium is a proliferation concern. There are various options to degrade the plutonium produced in the breeder blanket. The obvious one is to blend the blanket plutonium with those produced from the reactor core during reprocessing. Other options try to prevent the generation of good quality plutonium (Pu). The Protected Plutonium Production (P{sup 3}) Project proposed by Tokyo Institute of Technology (TIT)1,2,3 advocates the doping of certain amount of neptunium (Np), or americium (Am) in fresh blanket fuel for irradiation. The increased production of {sup 238}Pu, {sup 240}Pu and {sup 242}Pu by neutron capture in {sup 237}Np and Am would degrade the blanket plutonium. However, as {sup 237}Np is a controlled material according to IAEA, its use as doping material in fresh blanket fuel presents a concern for nuclear proliferation. In addition, the fabrication of fresh blanket fuel with inclusion of americium would be complicated due to the emission of intense low-energy gamma radiation from {sup 241}Am. Am is normally accompanied by Cm since the separation of those 2 elements is very difficult. Fuel containing both Am and Cm may make Safeguards measurement difficult. A variation would be doping the fresh blanket fuel with minor actinide (e.g., a group of neptunium, americium, and curium), or with separated reactor-grade (RG) plutonium. The drawback of such schemes would be the need for glove boxes in fresh blanket fuel fabrication. It is possible to fuel the breeder blankets with recycled (reprocessed) uranium oxide. The recycled uranium, recovered from reprocessing, contains {sup 236}U, which when irradiated in the blanket would

  1. Large scale breeder reactor pump dynamic analyses

    International Nuclear Information System (INIS)

    The lateral natural frequency and vibration response analyses of the Large Scale Breeder Reactor (LSBR) primary pump were performed as part of the total dynamic analysis effort to obtain the fabrication release. The special features of pump modeling are outlined in this paper. The analysis clearly demonstrates the method of increasing the system natural frequency by reducing the generalized mass without significantly changing the generalized stiffness of the structure. Also, a method of computing the maximum relative and absolute steady state responses and associated phase angles at given locations is provided. This type of information is very helpful in generating response versus frequency and phase angle versus frequency plots

  2. Fast Breeder Test Reactor: 15 years of operating experience

    International Nuclear Information System (INIS)

    Fast Breeder Test Reactor (FBTR) is a 40 MWt/13.2 MWe sodium cooled, loop type, mixed carbide-fuelled reactor. Its main aim is to gain experience in the design, construction and operation of fast reactors including sodium systems and to serve as an irradiation facility for development of fuel and structural materials for future fast reactors. It achieved first criticality in Oct 1985 with Mark I core (70% PuC - 30% UC). Steam generator was put in service in Jan 1993 and power was raised to 10.5 MWt in Dec 1993. Turbine generator was synchronized to the grid in Jul 1997. The indigenously developed mixed carbide fuel has achieved a peak burn up of 88,000 MWd/t till now at a linear heat rating of 320 W/cm and reactor power of 13.4 MWt without any fuel-clad failure. The paper presents operating and decontamination experience, performance of fuel, steam generator and sodium circuits, certain unusual occurrences encountered by the plant and various improvements carried out in reactor systems to enhance plant availability. (author)

  3. Liquid metal tribology in fast breeder reactors

    International Nuclear Information System (INIS)

    Liquid Metal Cooled Fast Breeder Reactors (LMFBR) require mechanisms operating in various sodium liquid and sodium vapor environments for extended periods of time up to temperatures of 900 K under different chemical properties of the fluid. The design of tribological systems in those reactors cannot be based on data and past experience of so-called conventional tribology. Although basic tribological phenomena and their scientific interpretation apply in this field, operating conditions specific to nuclear reactors and prevailing especially in the nuclear part of such facilities pose special problems. Therefore, in the framework of the R and D-program accompanying the construction phase of SNR 300 experiments were carried out to provide data and knowledge necessary for the lay-out of friction systems between mating surfaces of contacting components. Initially, screening tests isolated material pairs with good slipping properties and maximum wear resistance. Those materials were subjected to comprehensive parameter investigations. A multitude of laboratory scale tests have been performed under largely reactor specific conditions. Unusual superimpositions of parameters were analyzed and separated to find their individual influence on the friction process. The results of these experiments were made available to the reactor industry as well as to factories producing special tribo-materials. (orig.)

  4. Liquid metal cooled fast breeder nuclear reactor constructions

    International Nuclear Information System (INIS)

    A description is given of a liquid metal cooled fast breeder nuclear reactor construction of the pool kind in which the primary vessel incorporates an annular yoke fabricated from arcuate segments. The yoke is suspended from the roof structure of the vault by a first annular series of tie straps arranged outside the primary vessel whilst a strongback on which the fuel assembly sits inside the primary vessel is supported from the yoke by a second series of tie straps. The yoke has upwardly and downwardly extending legs which are extended by upper and lower strakes respectively of the primary vessel. (U.K.)

  5. Fast breeder reactors: experience and trends. V. 1

    International Nuclear Information System (INIS)

    The IAEA Symposium on ''Fast Breeder Reactors: Experience and Future Trends'' was held, at the invitation of the Government of France, in Lyons, France, on 22-26 July 1985. It was hosted by the French Commissariat a l'energie atomique and Electricite de France. The purpose of the Symposium was to review the experience gained so far in the field of LMFBRs, taking into account the constructional, operational, technological, economic and fuel cycle aspects, and to consider the developmental trends as well as the international co-operation in fast breeder reactor design and utilization. The Symposium presentations were divided into sessions devoted to the following topics: Experience of LMFBR construction and operation and resultant development strategies (6 papers); LMFBR plant startup and commissioning tests and general behaviour (8 papers); Core performance experience for high burnup and core design trends (8 papers); Experience and trends in the LMFBR fuel cycle (4 papers); Core design and behaviour (3 papers); Fuels and materials (7 papers). A separate abstract was prepared for each of these papers

  6. Reflections on the introduction of fast breeder reactors in the DeBeNeLux states

    International Nuclear Information System (INIS)

    This report gives a survey of the impact of introducing sodium-cooled fast breeder reactors in the Federal Republic of Germany and the BeNeLux countries (DeBeNeLux region). The supply situation with respect to electric and thermal energy is studied in particular, together with aspects of economy and environmental impact. The potential and consequences of a breeder economy, the present status and future r+d work are discussed. In addition to sodium-cooled fast breeder reactors with oxide or carbide fuel, alternative solutions are touched: 1) light water and high temperature reactors, 2) helium-cooled fast breeder reactors, 3) geothermal energy, solar energy and fusion energy. (orig.)

  7. Fuel Cycle Economics of Fast Breeders with Plutonium

    International Nuclear Information System (INIS)

    Pu-fuelled fast breeder systems are characterized by their attractive fuel cycle economics. Basically, the economics are influenced by a number of reactor parameters like fissile material rating, fuel bum-up, breeding ratio and thermal efficiency, on the one hand, and by a number of economic parameters like the plutonium price, the interest rate and the fabrication and reprocessing costs on the other. To a certain extent, the two sets of parameters are interdependent and the cost parameters are influenced by the existing nuclear industry as well. In the present paper it is shown, with the help of a number of specific examples, that the fuel cycle of Pu fast breeders is not a static and isolated property of the reactor but is dynamic in nature. Depending on the cost situation and other conditions, the fuel cycle can always be optimized anew to fit into the existing overall economics. A high Pu price, for example, requires a high fissile rating or a high breeding ratio, whereas, if the Pu price falls, neither a high rating nor a high breeding ratio is necessary to keep the fuel cycle costs low. The influence of fabrication costs may be regulated to some extent by varying the burn-up. The effect of reprocessing costs may be made comparatively insignificant provided reprocessing can be carried out in large centrally located multi-purpose plants for converter elements. Because of the high flexibility of the fuel cycle of Pu fast breeders, the attractiveness of their fuel cycle economics can be retained under a wide range of competitive conditions. (author)

  8. A contribution to the analysis of the thermal behaviour of Fast Breeder fuel rods with UO{sub 2}-PuO{sub 2} fuel; Contribucion al analisis del comportamiento termico de las barras combustibles de UO{sub 2}-PuO{sub 2} de los reactores rapidos

    Energy Technology Data Exchange (ETDEWEB)

    Lopez Jimenez, J.; Elbel, H.

    1977-07-01

    The fuel of Fast Breeder Reactors which consists of Uranium and Plutonium dioxide is mainly characterized by the amount and distribution of void volume and Plutonium and the amount of oxygen. Irradiation experiments carried out with this fuel have shown that initial structure of the fuel pellet is subjected to large changes during operation. These are consequences of the radial and axial temperature gradients within the fuel rods. (Author) 54 refs.

  9. Water chemistry of breeder reactor steam generators

    International Nuclear Information System (INIS)

    The water quality requirements will be described for breeder reactor steam generators, as well as specifications for balance of plant protection. Water chemistry details will be discussed for the following power plant conditions: feedwater and recirculation water at above and below 5% plant power, refueling or standby, makeup water, and wet layup. Experimental data will be presented from tests which included a departure from nucleate boiling experiment, the Few Tube Test, with a seven tube evaporator and three tube superheater, and a verification of control and on-line measurement of sodium ion in the ppB range. Sampling and instrumentation requirements to insure adherence to the specified water quality will be described. Evaporator cleaning criteria and data from laboratory testing of chemical cleaning solutions with emphasis on flow, chemical composition, and temperature will be discussed

  10. CANDU: The fuel conserving reactor

    International Nuclear Information System (INIS)

    Because of their high neutron economy and unique design features, CANDU heavy water moderated reactors are the only established commercial reactors able to use directly low fissile content fuels such as natural uranium or uranium recovered from spent light water reactor fuel (RU). These features also help them to achieve the highest fuel utilization of all commercially available reactors, whether the fuel is based on natural uranium or mixed oxides of plutonium, uranium or thorium. As nuclear capacity growth increases demands on the world's finite uranium resources, AECL envisages near term use in CANDU reactors of a fuel incorporating RU and fuels containing thorium, with either plutonium or low enriched uranium (LEU) as the fissile 'driver' fuel. In the long term, AECL proposes the use of future 'Generation X' CANDU reactors with enhanced neutron economy to achieve a near-Self-Sufficient Equilibrium Thorium (SSET) fuel cycle. This CANDU SSET would have a conversion ratio of unity and be able to produce power indefinitely, with the need for little additional fissile material once equilibrium is reached (the amount of 233U needed in the fresh fuel is the same as is present in the discharged fuel, including processing losses.) This would also enable a CANDU-Fast Breeder Reactor (FBR) synergism that would allow each fuel-generating, though expensive, FBR to supply the initial fissile requirements of several less-expensive, CANDU SSET reactors operating on the thorium cycle. The closer the approach to an SSET that CANDUs can achieve, the higher the ratio of CANDUs to breeders in an economically optimized reactor fleet. CANDU reactors thereby become natural partners of both light water-cooled thermal reactors and fast breeder reactors, in both cases making optimum use of their spent fuel components and enhancing the overall sustainability of nuclear power. (authors)

  11. Calculations of two-phase flows in the liquid metal cooled fast breeder reactors

    International Nuclear Information System (INIS)

    Mathematical models used for the safety analysis of liquid metal cooled fast breeder reactors are considered. Models, taking into account sodium boiling in reactor channels (one-dimensional and many-dimensional approaches), fuel cladding melting, and movement of molten materials during loss of coolant, accidents are described

  12. On the development of fast breeder reactors and the use of thorium in Brazil

    International Nuclear Information System (INIS)

    This work presents a discussion on the possibility of construction of fast breeder reactors in Brazil. It is specially concerned with the use of thorium which is abundant in our country. The main advantages of this projects are: develop fuel and reactor technology in Brazil, increase thorium research, demonstrate the safety of LMFBR and promote its public acceptance. (A.C.A.S.)

  13. Trace metal assay of fast breeder test reactor fuel using D.C. arc and plasma emission spectrometry

    International Nuclear Information System (INIS)

    This report describes the methods developed and used for the trace metal assay of the first charge of the FBTR fuel using a glove-box adapted direct reading emission spectrometer. The group of medium and highly volatile elements are determined in (U,Pu)C fuel samples by d.c. arc carrier distillation technique while the group of lanthanide elements are determined by ICP excitation mode with prior chemical separation from the major matrix. The statistical treatment of the analytical data collected from the analysis of about one hundred samples has indicated good purity of samples and consistent and satisfactory performance of the direct reading spectrometer and associated systems during this period. (author)

  14. Research and development long term prospects for advanced thermal and fast breeder reactors and fuel cycle activities

    International Nuclear Information System (INIS)

    Despite the present slowing down of most of the national nuclear programmes due to various conjunctural causes, nuclear energy is presumably to have in term an increasing role worldwide for electricity production, both in industrialized countries already equipped with nuclear power stations and in countries which intent to enter the nuclear energy route in a short or mid-term perspective. Nuclear energy development will have notably to take into account the following requirements: from a strategic point of view, the major concern is to optimize the use of world available natural uranium appears abundant and inexpensive, it must not be forgotten that in a longer time perspective a growing scarcity of this material will occur unavoidably; from the technical and economical standpoints, future reactors will have to be more reliable, more available, and even more safe, with lower KWh generating costs, than units presently constructed or operated. The paper presents the approach which is followed by France to meet these goals. (author)

  15. Molten Salt Breeder Reactor Analysis Based on Unit Cell Model

    International Nuclear Information System (INIS)

    Contemporary computer codes like the MCNP6 or SCALE are only good for solving a fixed solid fuel reactor. However, due to the molten-salt fuel, MSR analysis needs some functions such as online reprocessing and refueling, and circulating fuel. J. J. Power of Oak Ridge National Laboratory (ORNL) suggested in 2013 a method for simulating the Molten Salt Breeder Reactor (MSBR) with SCALE, which does not support continuous material processing. In order to simulate MSR characteristics, the method proposes dividing a depletion time into short time intervals and batchwise reprocessing and refueling at each step. We are applying this method by using the MCNP6 and PYTHON and NEWT-TRITON-PYTHON and PYTHON code systems to MSBR. This paper contains various parameters to analyze the MSBR unit cell model such as the multiplication factor, breeding ratio, change of amount of fuel, amount of fuel feeding, and neutron flux distribution. The result of MCNP6 and NEWT module in SCALE show some difference in depletion analysis, but it still seems that they can be used to analyze MSBR. Using these two computer code system, it is possible to analyze various parameters for the MSBR unit cells such as the multiplication factor, breeding ratio, amount of material, total feeding, and neutron flux distribution. Furthermore, the two code systems will be able to be used for analyzing other MSR model or whole core models of MSR

  16. Status of fast breeder reactor development in the United States of America - April 1984

    International Nuclear Information System (INIS)

    The Breeder Technology program continues to produce viable information on fuel performance, nuclear systems technology, and power conversion technology. The unique testing capabilities design into the FFTF have resulted in well-validated materials and fuels irradiation information that has confirmed and extended previous data bases. Current directions for the research and development program are to improve the technology for power conversion systems, components, instrumentation, and materials technology to the point where cost reduction and reliability potentials are realized. Operation of the breeder test facility complex at the Hanford Engineering Development Laboratory (HEDL), the Energy Technology Engineering Center (ETEC), and the Argonne National Laboratory (ANL) continues to provide the experience base and test capability for the breeder R and D effort. International cooperation will be even more important in the future than in the past for several reasons. Significant new investments still have to be made in breeder R and D to improve designs, achieve economic competitiveness and to develop practical breeder fuel cycle capabilities. Progress can be accelerated, redundancies avoided, and economics achieved if nations coordinate their programs, and where possible, divide up the work. In addition, there is clear mutual benefit in encouraging the countries involved in breeder development to harmonize standards and regulations related to safety. It is also important that the advanced nations work together closely in assuring that adequate international safeguards, export controls, and national physical security measures keep pace with breeder reactor and fuel cycle developments

  17. Current status of development of Demonstration Fast Breeder Reactor and prospect of FBR commercialization

    International Nuclear Information System (INIS)

    The Demonstration Fast Breeder Reactor (DFBR) is the next step of FBR development following the prototype fast breeder reactor 'MONJU'. The DFBR is now under development by The Japan Atomic Power Co. (JAPC) under the sponsorship of 9 Japanese electric power companies and Electric Power Development Co., Ltd. The JAPC has been performing the design study and R and D for DFBR in cooperation with Power Reactor and Nuclear Fuel Development Corp. (PNC), Central Research Institute of Electric Power Industry (CRIEPI) and Japan Atomic Energy Research Institute (JAERI). This report describes the prospect of FBR commercialization and the current status of new technology for DFBR and innovative technology FBR commercialization. (author)

  18. The design of the Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    India has a moderate uranium reserve and a large thorium reserve. The primary energy resource for electricity generation in the country is coal. The potential of other resources like gas, oil, wind, solar and biomass is very limited. The only viable and sustainable resource is the nuclear energy. Presently, Pressurised Heavy Water Reactors utilizing natural uranium are in operation/under construction and the plutonium generated from these reactors will be multiplied through breeding in fast breeder reactors. The successful construction, commissioning and operation of Fast Breeder Test Reactor at Kalpakkam has given confidence to embark on the construction of the Prototype Fast Breeder Reactor (PFBR). This paper describes the salient design features of PFBR including the design of the reactor core, reactor assembly, main heat transport systems, component handling, steam water system, electrical power systems, instrumentation and control, plant layout, safety and research and development

  19. Reactor physics and reactor strategy investigations into the fissionable material economy of the thorium and uranium cycle in fast breeder reactors and high temperature reactors

    International Nuclear Information System (INIS)

    In this work the properties governing the fissionable material economy of the uranium and thorium cycles are investigated for the advanced reactor types currently under development - the fast breeder reactor (FBR) and the high temperature reactor (HTR) - from the point of view of the optimum utilization of the available nuclear fuel reserves and the continuance of supply of these reserves. For this purpose, the two reactor types are first of all considered individually and are subsequently discussed as a complementary overall system

  20. Fast-breeder-power reactor records in the INIS database

    International Nuclear Information System (INIS)

    This report presents a statistical analysis of more than 19,700 records of publications concerned with research and technology in the field of fast breeder power fission reactors which are included in the INIS Bibliographic Database for the period from 1970. to 1999. The main objectives of this bibliometric study were: to make an inventory of the fast breeder power reactor related records in the INIS Database; to provide statistics and scientific indicators for the INIS users, namely science managers, researchers, engineers, operators, scientific editors and publishers, decision-makers in the field of fast breeder power reactors related subjects; to extract other useful information from the INIS Bibliographic Database about articles published in fast breeder reactors research and technology. The quantitative data in this report are obtained for various properties of relevant INIS records such as year of publication, secondary subject categories, countries of publication, language, publication types, literary types, etc. (author)

  1. ORIGEN2 model and results for the Clinch River Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Croff, A G; Bjerke, M A

    1982-06-01

    Reactor physics calculations and literature information acquisition have led to the development of a Clinch River Breeder Reactor (CRBR) model for the ORIGEN2 computer code. The model is based on cross sections taken directly from physics codes. Details are presented concerning the physical description of the fuel assemblies, the fuel management scheme, irradiation parameters, and initial material compositions. The ORIGEN2 model for the CRBR has been implemented, resulting in the production of graphical and tabular characteristics (radioactivity, thermal power, and toxicity) of CRBR spent fuel, high-level waste, and fuel-assembly structural material waste as a function of decay time. Characteristics for pressurized water reactors (PWRs), commercial liquid-metal fast breeder reactors (LMFBRs), and the Fast Flux Test Facility (FFTF) have also been included in this report for comparison with the CRBR data.

  2. Exploding the myths about the fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Burns, S.

    1979-01-01

    This paper discusses the facts and figures about the effects of conservation policies, the benefits of the Clinch River Breeder Reactor demonstration plant, the feasibility of nuclear weapons manufacture from reactor-grade plutonium, diversion of plutonium from nuclear plants, radioactive waste disposal, and the toxicity of plutonium. The paper concludes that the U.S. is not proceeding with a high confidence strategy for breeder development because of a variety of false assumptions.

  3. Immediate relation of ING to fast breeder reactor programs

    International Nuclear Information System (INIS)

    The future large-scale use of nuclear energy is linked in the United States and other major countries to their fast breeder reactor development. Very serious basic problems have been discovered within the last two years, limiting the life in the high fast neutron flux at appropriate temperatures of materials, in particular of metals suitable for fuel cladding in sodium coolant. There is therefore a most urgent need for materials testing facilities under controlled conditions of temperature and neutron flux at sufficiently high ratings to match or surpass those required in commercially competitive fast breeder reactors. None of the test facilities yet planned for 1976 or sooner in the western world appears to match these conditions. The problem is mainly the difficulty of providing the high neutron flux effectively continuously. The spallation reaction in heavy elements was chosen as the basis of ING - the intense neutron generator, because it is the only known reaction that promises a fast neutron source density that is higher than can be controlled from the fission process. It is suggested that several countries will wish to consider urgently whether they should also explore the spallation reaction for the purpose of a fast neutron irradiation test facility. In view of the discontinuance of the ING project in Canada a favourable opportunity will exist over the next few months 10 obtain from Canada by direct personal contact details of the significant study that has been carried on for ING over the last five years. In the event that satisfactory materials are established within the lifetime of the spallation facilities they may continue to be used for the production of selected isotopes more profitably produced in high neutron fluxes. The facilities may be also used for the desirable preirradiation of thorium reactor fuel. The other research purposes planned for ING could also be served. (author)

  4. Experience with the generating plant at fast breeder test reactor

    International Nuclear Information System (INIS)

    Fast Breeder Test Reactor (FBTR) is a 40 MWth/13.2 MW(e) sodium cooled, loop type, mixed carbide-fuelled reactor. Its main aim is to gain experience in the design, construction and operation of fast reactors including sodium systems and generating systems and to serve as an irradiation facility for development of fuel and structural materials for future fast reactors. It achieved first criticality in Oct.1985 with Mark-I core (70 % PUC-30 % UC). FBTR heat transport system consists of two primary sodium loops, two secondary sodium loops and one common tertiary steam and water circuit. Heat generated in the reactor core is transported to the tertiary loop by primary and secondary sodium loops. The steam water system mainly consists of a once through steam generator, which produces super heated steam at a pressure of 120 bars and temperature of 480 degC, feed water system and condensate system. The steam produced is supplied to a condensing turbine. The turbine in turn is coupled to an alternator. The steam generator was put in service in Jan.1993 and turbine generator was synchronized to the grid in July 1997. The paper presents operating experience with generating plant consisting of steam water circuit, condensing turbine and its associated systems and the alternator, various modifications carried out to improve system reliability and availability and certain incidents taken place in the generating plant. (author)

  5. An option for the Brazilian nuclear project: necessity of fast breeder reactors and core design for an experimental fast reactor

    International Nuclear Information System (INIS)

    Aiming to assure the continued utilization of fission energy, the development of fast breeder reactors (FBRs) is a necessity. Binary fueled LMFBRs are proposed, as the best type for the Brazilian nuclear system in the future. The inherent safety characteristics are superior to current fast breeder reactors and an efficient utilization of thorium can be realized. The construction and operation of an experimental fast reactor is the first step and a basic tool for the development of FBRs technologies. A serie of core design for an 90 MW FBR is studied and the possible options and sizes of the main parameters are identified. (E.G.)

  6. Breeder Reprocessing Engineering Test

    Energy Technology Data Exchange (ETDEWEB)

    Burgess, C.A.; Meacham, S.A.

    1984-01-01

    The Breeder Reprocessing Engineering Test (BRET) is a developmental activity of the US Department of Energy to demonstrate breeder fuel reprocessing technology while closing the fuel cycle for the Fast Flux Test Facility (FFTF). It will be installed in the existing Fuels and Materials Examination Facility (FMEF) at the Hanford Site near Richland, Washington, The major objectives of BRET are: (1) close the US breeder fuel cycle; (2) develop and demonstrate reprocessing technology and systems for breeder fuel; (3) provide an integrated test of breeder reactor fuel cycle technology - rprocessing, safeguards, and waste management. BRET is a joint effort between the Westinghouse Hanford Company and Oak Ridge National Laboratory. 3 references, 2 figures.

  7. Accident analysis of heavy water cooled thorium breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yulianti, Yanti [Department of Physics, University of Lampung Jl. Sumantri Brojonegoro No.1 Bandar Lampung, Indonesia Email: y-yanti@unila.ac.id (Indonesia); Su’ud, Zaki [Department of Physics, Bandung Institute of Technology Jl. Ganesha 10 Bandung, Indonesia Email: szaki@fi.itb.ac.id (Indonesia); Takaki, Naoyuki [Department of Nuclear Safety Engineering Cooperative Major in Nuclear Energy (Graduate School) 1-28-1 Tamazutsumi,Setagayaku, Tokyo158-8557, Japan Email: ntakaki@tcu.ac.jp (Japan)

    2015-04-16

    Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k, and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The

  8. Accident analysis of heavy water cooled thorium breeder reactor

    International Nuclear Information System (INIS)

    Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k, and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The

  9. Accident analysis of heavy water cooled thorium breeder reactor

    Science.gov (United States)

    Yulianti, Yanti; Su'ud, Zaki; Takaki, Naoyuki

    2015-04-01

    Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The power reactor has a peak value

  10. Status of liquid metal cooled fast breeder reactors

    International Nuclear Information System (INIS)

    This document represents a compilation of the information on the status of fast breeder reactor development. It is intended to provide complete and authoritative information for academic, energy, industrial and planning organizations in the IAEA Member States. The Report also provides extended reference and bibliography lists. A summarized overview of the national programmes of LMFBR development is given in Chapter II. Chapter III on LMFBR experience provides a brief description and purpose of all fast reactors - experimental, demonstration and commercial size - that have been or are planned for construction and operation. Fast reactor physics is dealt with in Chapter IV. Besides the basic facts and definitions of neutronics and the compilation and measurement of nuclear data, a broad range of the calculation methods, codes, and the state of the art is described. In Chapter V, fuels and materials are described. The emphasis is on the design and development experience gained with mixed oxide fuel pins and subassemblies. Structural materials, blanket elements and absorber materials are also discussed. Chaper VI presents a broad overview of the technical and engineering aspects of LMFBR power plants. LMFBR core design is described in detail, followed by the components of the main heat transport system, the refuelling equipment, and auxiliary systems. Chapter VII on safety is a compilation of the current safety design concepts of LMFBRs and new trends in safety criteria and safety goals. The chapter concludes with risk analyses of LMFBR technology. In Chapter VIII, the systems approach has been emphasized in the consideration of the whole LMFBR fuel cycle. Special emphasis is placed on safeguards aspects and the environmental impact of the LMFBR fuel cycle. Chapter IX describes deployment considerations of LMFBRs. Special emphasis is placed on economic aspects of the LMFBR power plant and its related fuel cycle. Finally, Chapter X provides an overall summary and a

  11. Nuclear Engineering Computer Modules, Thermal-Hydraulics, TH-2: Liquid Metal Fast Breeder Reactors.

    Science.gov (United States)

    Reihman, Thomas C.

    This learning module is concerned with the temperature field, the heat transfer rates, and the coolant pressure drop in typical liquid metal fast breeder reactor (LMFBR) fuel assemblies. As in all of the modules of this series, emphasis is placed on developing the theory and demonstrating the use with a simplified model. The heart of the module is…

  12. A five years experience of pulse columns extraction cycles for the reprocessing of fast breeder reactor fuels at the Marcoule pilot plant (SAP)

    International Nuclear Information System (INIS)

    The reprocessing of Phenix fast breeder reactor started at the MARCOULE PILOT PLANT in 1977 with the enriched UO2 first core (2.3 tons U) followed by several campaigns of UO2-PuO2 Phenix-core II (6.5 tons U-Pu). After a short description of the Pilot Plant, characteristics of the pulse columns extraction flow-sheets are presented. Pulse columns are used for extraction and scrubbing of uranium and plutonium and for uranium backwashing whilst plutonium stripping and U-Pu partition are carried out in mixer settlers with HAN and in-line electrolytic U IV generation. Performances of pulsed columns including recovery yields and decontamination factors are discussed: they show a good β γ decontamination can be reached with two cycles and partition carried out at the second cycle. (author)

  13. A five years experience of pulse columns extraction cycles for the reprocessing of fast breeder reactor fuels at the Marcoule pilot plant (SAP)

    International Nuclear Information System (INIS)

    The reprocessing of Phenix fast breeder reactor started at the Marcoule Pilot Plant in 1977 with the enriched UO2 first core (2.3 tons U) followed by several campaigns of UO2-PuO2 Phenix-core II (6.5 tons U-Pu). After a short description of the Pilot Plant, characteristics of the pulse columns extraction flow-sheets are presented. Pulse columns are used for extraction and scrubbing of uranium and plutonium and for uranium backwashing whilst plutonium stripping and U-Pu partition are carried out in mixer settlers with HAN and in-line electrolytic U IV generation. Performances of pulsed columns including recovery yields and decontamination factors are discussed: they show a good β γ decontamination can be reached with two cycles and partition carried out at the second cycle

  14. Fast reactor fuel cycle facility

    International Nuclear Information System (INIS)

    An integrated fuel cycle facility named Fast Reactor Fuel Cycle Facility (FRFCF) is planned to be set up at Kalpakkam to close the fuel cycle of the Prototype Fast Breeder Reactor (PFBR) that is already under construction there. FRFCF is the first project of its kind in India. Closure of fuel cycle of PFBR will be a significant milestone of the second stage of nuclear power programme of the Department of Atomic Energy. The facility would be ready for operation in 2014. Design work and safety review of FRFCF are presently in progress. (author)

  15. Molten Salt Breeder Reactor Analysis Methods

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jinsu; Jeong, Yongjin; Lee, Deokjung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2015-05-15

    Utilizing the uranium-thorium fuel cycle shows considerable potential for the possibility of MSR. The concept of MSBR should be revised because of molten salt reactor's advantage such as outstanding neutron economy, possibility of continuous online reprocessing and refueling, a high level of inherent safety, and economic benefit by keeping off the fuel fabrication process. For the development of MSR research, this paper provides the MSBR single-cell, two-cell and whole core model for computer code input, and several calculation results including depletion calculation of each models. The calculations are carried out by using MCNP6, a Monte Carlo computer code, which has CINDER90 for depletion calculation using ENDF-VII nuclear data. From the calculation results of various reactor design parameters, the temperature coefficients are all negative at the initial state and MTC becomes positive at the equilibrium state. From the results of core rod worth, the graphite control rod alone cannot makes the core subcritical at initial state. But the equilibrium state, the core can be made subcritical state only by graphite control rods. Through the comparison of the results of each models, the two-cell method can represent the MSBR core model more accurately with a little more computational resources than the single-cell method. Many of the thermal spectrum MSR have adopted a multi-region single-fluid strategy.

  16. Shutdown and Closure of the Experimental Breeder Reactor - II

    International Nuclear Information System (INIS)

    The Department of Energy mandated the termination of the Integral Fast Reactor (IFR) Program, effective October 1, 1994. To comply with this decision, Argonne National Laboratory-West (ANL-W) prepared a plan providing detailed requirements to maintain the Experimental Breeder Reactor - II (EBR-II) in a radiologically and industrially safe condition, including removal of all irradiated fuel assemblies from the reactor plant, and removal and stabilization of the primary and secondary sodium, a liquid metal used to transfer heat within the reactor plant. The EBR-II is a pool-type reactor. The primary system contained approximately 325 m3 (86,000 gallons) of sodium and the secondary system contained 50 m3 (13,000 gallons). In order to properly dispose of the sodium in compliance with the Resource Conservation and Recovery Act (RCRA), a facility was built to react the sodium to a solid sodium hydroxide monolith for burial as a low level waste in a land disposal facility. Deactivation of a liquid metal fast breeder reactor (LMFBR) presents unique concerns. Residual amounts of sodium remaining in circuits and components must be passivated, inerted, or removed to preclude future concerns with sodium-air reactions that could generate potentially explosive mixtures of hydrogen and leave corrosive compounds. The passivation process being implemented utilizes a moist carbon dioxide gas that generates a passive layer of sodium carbonate/sodium bicarbonate over any quantities of residual sodium. Tests being conducted will determine the maximum depths of sodium that can be reacted using this method, defining the amount that must be dealt with later to achieve RCRA clean closure. Deactivation of the EBR-II complex is on schedule for a March, 2002, completion. Each system associated with EBR-II has an associated lay-up plan defining the system end state, as well as instructions for achieving the lay-up condition. A goal of system-by-system lay-up is to minimize surveillance and

  17. Shutdown and closure of the experimental breeder reactor - II

    International Nuclear Information System (INIS)

    The Department of Energy mandated the termination of the Integral Fast Reactor (IFR) Program, effective October 1, 1994. To comply with this decision, Argonne National Laboratory-West (ANL-W) prepared a plan providing detailed requirements to maintain the Experimental Breeder Reactor-II (EBR-II) in a radiologically and industrially safe condition, including removal of all irradiated fuel assemblies from the reactor plant, and removal and stabilization of the primary and secondary sodium, a liquid metal used to transfer heat within the reactor plant. The EBR-II is a pool-type reactor. The primary system contained approximately 325 m3 (86,000 gallons) of sodium and the secondary system contained 50 m3 (13,000 gallons). In order to properly dispose of the sodium in compliance with the Resource Conservation and Recovery Act (RCRA), a facility was built to react the sodium to a solid sodium hydroxide monolith for burial as a low level waste in a land disposal facility. Deactivation of a liquid metal fast breeder reactor (LMFBR) presents unique concerns. Residual amounts of sodium remaining in circuits and components must be passivated, inerted, or removed to preclude future concerns with sodium-air reactions that could generate potentially explosive mixtures of hydrogen and leave corrosive compounds. The passivation process being implemented utilizes a moist carbon dioxide gas that generates a passive layer of sodium carbonate/sodium bicarbonate over any quantities of residual sodium. Tests being conducted will determine the maximum depths of sodium that can be reacted using this method, defining the amount that must be dealt with later to achieve RCRA clean closure. Deactivation of the EBR-II complex is on schedule for a March, 2002, completion. Each system associated with EBR-II has an associated layup plan defining the system end state, as well as instructions for achieving the layup condition. A goal of system-by-system layup is to minimize surveillance and

  18. Design evaluation system for class 1 component of fast breeder reactor plants

    International Nuclear Information System (INIS)

    The development of a new type of nuclear power plant called Fast Breeder Reactor has been greatly promoted of late by the Power Reactor and Nuclear Fuel Development Corporation (PNC) and others in hopes of replacing Light Water Reactors so far prevailing in Japan. Fast Breeder Reactor, unlike Light Water Reactor, is subjected to elevated temperature within the creep temperature range for long duration, thus requiring higher structural standards for reliability as well as for safety. In this connection, PNC has been conducting many years' research and development to establish reliable design methods based on an advanced analysis taking into consideration elevated temperature properties of materials, and finally worked out Structural Design Guide for Class 1 Components of the prototype of Fast Breeder Reactor in elevated temperature service. The POST-DS system in this paper has been developed as an design evaluation system based on the above design guide, by Mitsui Engineering and Shipbuilding Co., Ltd. since 1979 in accordance with a commission given by PNC. Using the results of the heat transfer analysis and stress analysis for Class 1 Components of Fast Breeder Reactor, this system can evaluate the following factors. 1) Primary stress limit, 2) Strain limit, 3) Creep Fatigue damage. (author)

  19. BREEDER: a microcomputer program for financial analysis of a large-scale prototype breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Giese, R.F.

    1984-04-01

    This report describes a microcomputer-based, single-project financial analysis program: BREEDER. BREEDER is a user-friendly model designed to facilitate frequent and rapid analyses of the financial implications associated with alternative design and financing strategies for electric generating plants and large-scale prototype breeder (LSPB) reactors in particular. The model has proved to be a useful tool in establishing cost goals for LSPB reactors. The program is available on floppy disks for use on an IBM personal computer (or IBM look-a-like) running under PC-DOS or a Kaypro II transportable computer running under CP/M (and many other CP/M machines). The report documents version 1.5 of BREEDER and contains a user's guide. The report also includes a general overview of BREEDER, a summary of hardware requirements, a definition of all required program inputs, a description of all algorithms used in performing the construction-period and operation-period analyses, and a summary of all available reports. The appendixes contain a complete source-code listing, a cross-reference table, a sample interactive session, several sample runs, and additional documentation of the net-equity program option.

  20. BREEDER: a microcomputer program for financial analysis of a large-scale prototype breeder reactor

    International Nuclear Information System (INIS)

    This report describes a microcomputer-based, single-project financial analysis program: BREEDER. BREEDER is a user-friendly model designed to facilitate frequent and rapid analyses of the financial implications associated with alternative design and financing strategies for electric generating plants and large-scale prototype breeder (LSPB) reactors in particular. The model has proved to be a useful tool in establishing cost goals for LSPB reactors. The program is available on floppy disks for use on an IBM personal computer (or IBM look-a-like) running under PC-DOS or a Kaypro II transportable computer running under CP/M (and many other CP/M machines). The report documents version 1.5 of BREEDER and contains a user's guide. The report also includes a general overview of BREEDER, a summary of hardware requirements, a definition of all required program inputs, a description of all algorithms used in performing the construction-period and operation-period analyses, and a summary of all available reports. The appendixes contain a complete source-code listing, a cross-reference table, a sample interactive session, several sample runs, and additional documentation of the net-equity program option

  1. Manufacturing of prototype fast breeder reactor components: challenges and achievements

    International Nuclear Information System (INIS)

    In the presentation, three components of 500 MWe Prototype Fast Breeder Reactor (PFBR), viz. grid plate, roof slab and fuel handling systems, are focused, which have been responsible for the considerable delay of the project schedule. The manufacturing challenges of grid plate mainly originated from large number of sleeves resulting in higher self weight and hard facing of large diameter sleeves. Machining of large diameter plates and shell assembly to the required tight tolerances on dimensions, hard facing with nickel based cobalt free hard facing material on continuous, large diameter (6.7 m) annular tracks, heat treatment of large austenitic stainless steel parts at 1050℃ with controlled rates of cooling and heating together with control on temperature gradient across the parts, complex assembly of a large number of parts (∼14900) meeting the important requirements on verticality of sleeve assemblies (Ø0.1 mm) and delicate handling and transportation are truly challenging activities in the manufacturing technology. In case of roof slab, complex manufacturing process, especially welding between the shell and stiffeners caused lamellar tearing problems and extensive testing time. Inclined fuel transfer machine, multiple repairs, heavy weight and testing strategy resulted in long manufacturing and testing time. Some general lessons learnt are also brought out in this presentation. Technology development prior to start of construction is essential for long delivery components. Judicious choice of tolerances, number and location of welds and inspections has to be made. Robust criteria need to be applied for the acceptance of manufacturing deviations and material compositions. Indigenous materials should be used after qualifications of manufacturing process of direct relevance apart from routine standards. From the rich experience gained through the manufacture and erection of reactor assembly components of PFBR, important guidelines and approaches were derived

  2. Applicability of three dimensional diffusion theory programmes based on coarse mesh methods to calculating nuclear characteristics of fast breeder reactors

    International Nuclear Information System (INIS)

    Hexagonal coarse mesh methods in three dimensional diffusion theory programme have been examined for calculating in detail nuclear characteristics of fast breeder reactors composed of hexagonal fuel assemblies, comparing with more accurate triangular fine mesh method. The fast breeder reactors considered here are LMFBRs with different core configurations including heterogeneous core and GCFRs in different burnup states. The nuclear characteristics investigated in the comparative study are effective multiplication factor, power and neutron flux distributions, breeding ratio, reactivity effects and control rod reactivity worth. The comparative study indicates that the conventional coarse mesh method is not adeguate to detailed evaluation on nuclear characteristics of fast breeder reactors, and that the improved coarse mesh method developed by T. Takeda et al. is very useful for this purpose, though some problems exists in evaluation of power distribution and breeding ratio of the extremely composite fast breeder reactors, such as the radially heterogeneous core LMFBR. (author)

  3. Clinch River Breeder Reactor Plant Project: construction schedule

    International Nuclear Information System (INIS)

    The construction schedule for the Clinch River Breeder Reactor Plant and its evolution are described. The initial schedule basis, changes necessitated by the evaluation of the overall plant design, and constructability improvements that have been effected to assure adherence to the schedule are presented. The schedule structure and hierarchy are discussed, as are tools used to define, develop, and evaluate the schedule

  4. Symposium on key questions about the fast breeder reactor

    International Nuclear Information System (INIS)

    Except for several introductions on various aspects of the fast breeder reactor development this paper contains the full texts of the discussions held in the sub-groups panels on resp. technical matters, environment and health, society, politics and economics. The main issues of each discussion are summarized

  5. Tritium dynamics in fusion reactor solid breeder

    International Nuclear Information System (INIS)

    In the field of the NET research progrm, the chemical and diffusive processes involved in solid ceramic breeder materials have been analysed. A mathematical model describing the phenomena has been developed to obtain a quantitative evaluation for a first design approach. The data obtained by means of the above mentioned model are in good agreement with the data obtained by other research groups working in Europe and in United States. The computer codes BLANKET2, MC2, FWBC, have been developed to simulate the phenomena

  6. Status of national programmes on fast breeder reactors

    International Nuclear Information System (INIS)

    The twenty-second Annual Meeting of the International Working Group on Fast Reactors took place in Vienna, 18-21 April 1989. Nineteen representatives from twelve Member States and International Organizations attended the Meeting. This publication is a collection of presentations in which the participants reported the status of their national programmes on fast breeder reactors. A separate abstract was prepared for each of the twelve papers from this collections. Refs, figs, tabs and 1 graph

  7. Analysis of thorium and uranium fuel cycles in an iso-breeder lead fast reactor using extended-EQL3D procedure

    International Nuclear Information System (INIS)

    Highlights: ► Extension of EQL3D procedure to calculate radio-toxicity and decay heat. ► Characterization of uranium- and thorium-fueled LFR from BOL to equilibrium. ► Safety improvements for a LFR in a closed thorium cycle. ► Advantages of thorium-fueled LFR in terms of decay heat and radio-toxicity generation. ► Safety, decay heat and radio-toxicity concerns for a Th–Pu beginning-of-life core. - Abstract: Use of thorium in fast reactors has typically been considered as a secondary option, mainly thanks to a possible self-sustaining thorium cycle already in thermal reactors and due to the limited breeding capabilities compared to U–Pu in the fast neutron energy range. In recent years nuclear waste management has become more important, and the thorium option has been reconsidered for the claimed potential to burn transuranic waste and the lower build-up of hazardous isotopes in a closed cycle. To ascertain these claims and their limitations, the fuel cycle isotopic inventory, and associated waste radio-toxicity and decay heat, should be quantified and compared to the case of the uranium cycle using realistic core configurations, with complete recycle of all the actinides. Since the transition from uranium to thorium fuel cycles will likely involve a transuranic burning phase, this transition and the challenges that the evolving fuel actinide composition presents, for instance on reactor feedback parameters, should also be analyzed. In the present paper, these issues are investigated based on core physics analysis of the Lead-cooled Fast Reactor ELSY, performed with the fast reactor ERANOS code and the EQL3D procedure allowing full-core characterization of the equilibrium cycle and the transition cycles. In order to compute radio-toxicity and decay heat, EQL3D has been extended by developing a new module, which has been assessed against ORIGEN-S and is presented here. The capability of the EQL3D procedure to treat full-core 3D geometries allowed to

  8. Report to the Congress: liquid metal fast breeder reactor program--past, present, and future, Energy Research and Development Administration

    International Nuclear Information System (INIS)

    The past, present, and future of the liquid metal fast breeder reactor (LMFBR) program, the Nation's highest priority energy program, are studied. ERDA anticipates that the operation of the first large commercial breeder will start in 1987, and that 186 commercial-size breeders will be in operation by the year 2000. The breeder program is made up of six major areas, each dealing with an important element of technology: reactor physics; fuels and materials; fuel recycle; safety; component development; plant experience; and facilities used in the LMFBR program. ERDA is implementing a new system for administering, managing, and controlling the breeder program that will provide increased program visibility and control. Federal funding for breeder development was $168 million in FY 1971, accounting for 40% of the total Federal R and D energy budget; in FY 1976 Federal funding for this program will be $474 million, only 26% of total Federal funding for energy research. Besides Federal funds, over half a billion dollars have been or will be invested by industry over the next 5 to 10 years to develop the breeder and to build a demonstration plant. Five other nations--the United Kingdom, France, Japan, West Germany, and the Soviet Union--have a high priority national energy program for developing the LMFBR. These foreign breeder programs could contribute important data and information to the U.S. program

  9. Light-water breeder reactors: preliminary safety and environmental information document. Volume III

    International Nuclear Information System (INIS)

    Information is presented concerning prebreeder and breeder reactors based on light-water-breeder (LWBR) Type 1 modules; light-water backfit prebreeder supplying advanced breeder; light-water backfit prebreeder/seed-blanket breeder system; and light-water backfit low-gain converter using medium-enrichment uranium, supplying a light-water backfit high-gain converter

  10. Progress report on fast breeder reactor development in Japan

    International Nuclear Information System (INIS)

    In the power increase performance test of the experimental fast reactor ''Joyo'', which was in progress since April, the first stage of the rated thermal output of 50 MW has been accomplished on July 5. Thereafter, the continuous opeation test at 50 MW for 100 hours was performed for the verification of its overall operational performance from August 13 to 16. The safety evaluation for power increase up to 75 MW and 100 MW, which was under way since September, last year, was completed, and the power increase was licensed on September 20. Concerning the design of the prototype fast breeder reactor ''Monju'', the studies on the specifications of the Construction Preliminary Design (2) have been finished. In respect of the analysis and preparation of materials for the Safety Licensing by the Committee, the developments of the analytical codes for rupture propagation in the heat transfer tubes of steam generators and for decay heat have been conducted. In the construction site surveys, the third geological structure survey and beach deformation survey have all ended, while the meteorological and seismic observations, the prediction of the diffusion of drained warm water, the survey of river flow, etc. are now under way. A report on the survey conducted on the construction site in Shiraki was received by the Fukui prefectural government in July, and the copies of a report on the assessment of environmental effect were submitted in August to both the national government and the Fukui prefectural government. The situations of progress of the research and development works on reactor physics, structural components, instrumentation and control, sodium technology, fuel materials, structural materials, safety and steam generators are reported. (Nakai, Y.)

  11. Summary report on reduction of environmental burden related to nuclear fuel cycle. Collaboration of feasibility study on commercialized fast breeder reactor cycle system between JNC and JAERI

    International Nuclear Information System (INIS)

    This report summaries the state of the art of collaborative study related to the reduction of environmental burden by applying partition and transmutation technologies between JNC and JAERI conducted on July 2000. The items discussed are listed below. (1) Evaluation Index for Reduction of Environmental Burden in the Feasibility Study. (2) Research and Development Plans for Partition and Transmutation Technologies. PT Technology Development in JNC. PT Technology Development in JAERI. (3) Information Exchange for Basic Technology Related to Partition and Transmutation. Fuel Manufacturing Technology. Nuclear Data. Reactor Physics. Separation Technology for Fission Products. (author)

  12. Elements for evaluation of fast breeder reactor's potential in Argentina

    International Nuclear Information System (INIS)

    Fast Breeder Reactors (FBR) main features are presented in a general form, including their physical principles, the history of their evolution, their relevant technological aspects and the basis for their comparison to other energy sources. This is completed with descriptions of typical reactors and a model of FBR penetration in the Argentine electrical network. It is recommended to form a multidisciplinary board to study which position should be taken with respect to this type of reactors. In the author's opinion a Research activity should be started and gradually increased for passing to Development activities after a short while. (Author)

  13. Challenges and achievements - Prototype Fast Breeder Reactor construction

    International Nuclear Information System (INIS)

    Prototype fast breeder reactor presently under construction poses several challenges in materials, design and construction. The civil structure and equipment are of very large size and complex in nature. This paper presents the features of the design and construction of the PFBR excavation, raft, civil structure of the nuclear island connected buildings and reactor vault. This paper also brings out the details of the large size equipment of special stainless steel and handling structure for their lifting and placement inside the reactor vault. The paper is divided into three parts viz. introduction, challenges and achievements during construction of civil structures and erection of large size components. (author)

  14. Achieving sustainability in fuel cycles with Th-fuelled thermal breeders

    International Nuclear Information System (INIS)

    This paper outlines the path to reach sustainable fuel cycles with Th-fuelled thermal breeder reactors. To achieve a successful synergy between radio-toxic waste transmutation from water cooled reactors and breeding of the new fuel in the Th-233U cycle, fast reactors (FRs) are applied to transmute minor actinides (MAs). It is shown that near-term 233U breeding in PWR and FR cores is feasible. Over 100 kg of 233U could be produced annually in PWR cores with 30% of sub-assemblies containing ThO2, which means that one 850 MWe pressurized heavy water moderated thermal self-breeder could be started every 16 years. A slightly higher figure, over 120 kg/yr of 233U, was obtained when 30% of UO2 pins in each sub-assembly were exchanged for ThO2 pins. Fast reactors employing (Th,TRU)O2 fuel produced up to 370 kg of 233U annually, which means that a new self-breeder could be started roughly every 4.4 years. At the same time, 78 kg of minor actinides are consumed annually. By the end of this century, PWRs and FRs could generate enough 233U to sustain an increase in nuclear power capacities to 1160 GWe, which represents more than a three-fold increase of installed nuclear capacities worldwide. By 2200, the amount of TRUs in the fuel cycle could also be decreased and stabilized. (authors)

  15. Status of fast breeder reactor development in the Federal Republic of Germany, Belgium and the Netherlands

    International Nuclear Information System (INIS)

    Problems of public acceptance of nuclear power have affected the development of fast breeder reactors in the Federal Republic of Germany. Besides cooperation with UK, USA and Japan, the most outstanding event in the field of international fast breeder cooperation was a set of the agreements between Germany and France. These agreements opened the possibility of joint fast breeder development by Germany together with Belgium and the Netherlands. Most activities on the site of Compact Sodium Cooled Nuclear Reactor KNK-II were concerned with commissioning of the plant and final construction work. Criticality was achieved in Oct. 1977 and low-power tests performed. This paper includes a description of the status of construction of SNR-300 reactor and the results of research and development programmes performed. These were concerned with fuel elements development and results of irradiation experiments; development of cladding materials and core element structural materials; interaction between fuel and cladding; sodium tests; development and verification of computer codes; experiments in fast critical assemblies; fast rector safety; core disruptive accidents; development of instrumentation; thermodynamics od fuel assemblies; fluid dynamics

  16. The Last Twenty Years of Experience with Fast Breeder Reactors: Lessons Learnt and Perspectives

    International Nuclear Information System (INIS)

    India has made significant achievements in the design and development of sodium cooled fast breeder reactors over the last twenty years. Attaining a maximum burnup of 165 GW.d/t for the plutonium-rich carbide fuel without any cladding failure, coupled with excellent performance of sodium components, including primary pumps, heat exchangers and steam generators over the last 24 years, reprocessing of carbide fuel with a burnup of 150 GW.d/t and engineering tests performed for validating the plant dynamics computer codes, are the achievements from the successful operation of the 40 MW(th) capacity loop type fast breeder test reactor. Indigenous design of the 500 MW(e) Prototype Fast Breeder Reactor (PFBR), executing high quality multidisciplinary R and D and successful manufacturing and erection of large dimensioned thin walled shell structures are the achievements in PFBR development. These achievements, apart from providing confidence in the PFBR project, are instrumental for the design of innovative future reactors. National and international collaboration established with R and D establishments and academic institutions would go a long way towards helping India to attain world leadership by 2020. (author)

  17. Safeguards in Prototype Fast Breeder Reactor Monju

    International Nuclear Information System (INIS)

    Conclusion: (1) The integrated safeguards approach (ISA) was applied to Monju in November 2009 after it was confirmed that fuel assemblies were able to be monitored in every fuel handling route using the safeguards equipments installed in Monju. (2) A series of design, development and improvement related to safeguards in Monju, which were started in about 1985 (on the Monju construction stage), was finished by the shift to the ISA. (3) The safeguards method in Monju is expected to be a future FBR safeguards model. (4) In future, the reliability of safeguards equipments will be confirmed by accumulating the handling achievements of the spent fuel assemblies etc

  18. World energy resources, demand and supply of energy, and the prospects for the fast breeder reactor

    International Nuclear Information System (INIS)

    In the past it was taken for granted that the prime role of fast breeder reactors was to complement light water reactors, mainly because of their similar and compatible fuel cycles. In particular, the plutonium converted in LWRs is most intelligently disposed of and used in FBRs. Evaluation of the time horizon of such reactor strategies generally extended only to the year 2000. It is important to realize, however, that the salient task in the breeder field after 2000 - besides electricity generation - will be to substitute for conventional ''cheap'' oil. Electricity today makes up only 10% to 12% of the total secondary energy, while liquids essentially command up to about 50%. Thus the future application of the FBR technology will have to be geared more to the production of a liquid secondary energy carrier than to electricity. A new yardstick for all these considerations is the strongly rising energy prices. They may double, for example, leading to an oil price of US 24/bbl. Under these circumstances it is prudent to generalize the scope for future fast breeders. The key element of such a new fast breeder strategy would be the production of hydrogen by electrolysis or thermolysis or a combination of both. For example, methanol synthesized from hydrogen and residual fossil fuels would thus become economically attractive. The FBR breeding gain, on the other hand, would be used for the continued supply of LWRs generating electricity. The paper identifies order-of-magnitude considerations most important for such a fast breeder application against a global energy demand scenario for the year 2030. (author)

  19. Status of national programmes on fast breeder reactors

    International Nuclear Information System (INIS)

    The present document contains information on the status of fast breeder reactor development and on worldwide activities in this advanced nuclear power technology during 1989 as reported at the 23rd meeting of the IWGFR in Vienna, April 1990. The publication is intended to provide information regarding the current status of LMFBR development in IAEA Member States. A separate abstract was prepared for each of the 11 papers presented by the participants of this meeting. Refs, figs and tabs

  20. Status of national programmes on fast breeder reactors

    International Nuclear Information System (INIS)

    The present document contains information on the status of fast breeder reactor development and on worldwide activities in this advanced nuclear power technology during 1990 as reported at the 24th meeting of the IWGFR in Tsuruga, Japan, 15-18 April 1991. The publication is intended to provide information regarding the current status of LMFBR development in IAEA Member States and CEC. Figs and tabs

  1. The long-term future for civilian nuclear power generation in France: The case for breeder reactors. Breeder reactors: The physical and physical chemistry parameters, associate material thermodynamics and mechanical engineering: Novelties and issues

    International Nuclear Information System (INIS)

    The author gives a summary overview of the knowledge base acquired since the first breeder reactors became operational in the fifties. Neutron transport theory, thermal phenomena, reactor core cooling, various coolants used and envisioned for this function, fuel fabrication from separated materials, main equipment (pumps, valves, heat exchanges...) have now attained maturity, sufficient to implement sodium cooling circuits. However, the use of metallic sodium still raises certain severe questions in terms of safe handling and security considerations. The structural components, both inside the reactor core and outside (i.e. heat exchangers) are undergoing in-depth research so as to last longer. The fuel cycle, notably the re-fabrication of fuel elements and fertile elements, the case of transuranic elements, etc., call for studies into radiation induced phenomena, chemistry separation, separate or otherwise treatments for materials that have different radioactive, physical, thermodynamical, chemical and biological properties. The concerns that surround the definitive disposal of certain radioactive wastes could be qualitatively improved with respect to the pressurized water reactors (PWRs) in service today. Lastly, the author notes that breeder reactors eliminate the need for an isotope separation facility, and this constitutes a significant contribution to contain nuclear proliferation. France was in the forefront of nuclear breeder power generation science, technological research and also in the knowledge base related to breeder reactors. It is in the country's interest to pursue these efforts. (author)

  2. Power generation costs for alternate reactor fuel cycles

    International Nuclear Information System (INIS)

    The total electric generating costs at the power plant busbar are estimated for various nuclear reactor fuel cycles which may be considered for power generation in the future. The reactor systems include pressurized water reactors (PWR), heavy-water reactors (HWR), high-temperature gas cooled reactors (HTGR), liquid-metal fast breeder reactors (LMFBR), light-water pre-breeder and breeder reactors (LWPR, LWBR), and a fast mixed spectrum reactor (FMSR). Fuel cycles include once-through, uranium-only recycle, and full recycle of the uranium and plutonium in the spent fuel assemblies. The U3O8 price for economic transition from once-through LWR fuel cycles to both PWR recycle and LMFBR systems is estimated. Electric power generation costs were determined both for a reference set of unit cost parameters and for a range of uncertainty in these parameters. In addition, cost sensitivity parameters are provided so that independent estimations can be made for alternate cost assumptions

  3. Plutonium breeding in liquid-metal fast breeder reactors and light water reactors

    International Nuclear Information System (INIS)

    The possibilities of breeding in liquid-metal fast breeder reactors (LMFBRs) and light water reactors (LWRs) are compared in two ways. The feasibility of breeding has been demonstrated in the Phenix reactor with a measured gain of 0.14. The gain in Superphenix will amount to about0.20. The studies show that while maintaining the performance of commercial reactors their breeding gain can be further increased either by the concept of heterogeneous cores or by using carbide or nitride fuel (breeding gain about0.35). Recently, the old idea of breeding in advanced pressurized water reactors (PWRs) has been taken up again with the objective of attaining a gain of 0.05. Unfortunately, these objectives had to be limited to a conversion ratio of 0.9 for safety reasons, and it is not certain whether operation will be rewarding economically. The strategy of substituting PWRs is examined using the French example. By gradually introducing LMFBRs, the cumulated uranium supplies in France can be kept within reasonable limits, which means that they attain three to four times the home resources. This is not possible with advanced LWRs, which can be considered only as a possible backup solution for plutonium recycling into PWRs

  4. Neutron irradiation of candidate ceramic breeder materials of fusion reactors

    International Nuclear Information System (INIS)

    In the context of the European programs for the future fusion reactors, the Process Chemistry Department of ENEA, Casaccia Center (Rome), has been involved in preparing ceramic blanket materials as tritium breeders; a special consideration has been addressed to the nuclear characterization of LiAlO2 and Li2ZrO3. In this paper are reported neutron irradiation of ceramic specimens in TRIGA reactor and γ-spectrometric measurements for INAA purposes; and isothermal annealing of the irradiated samples and tritium extraction, by using an 'out of pile' system. (author) 3 refs.; 4 figs.; 4 tabs

  5. Reliability modeling of Clinch River breeder reactor electrical shutdown systems

    International Nuclear Information System (INIS)

    The initial simulation of the probabilistic properties of the Clinch River Breeder Reactor Plant (CRBRP) electrical shutdown systems is described. A model of the reliability (and availability) of the systems is presented utilizing Success State and continuous-time, discrete state Markov modeling techniques as significant elements of an overall reliability assessment process capable of demonstrating the achievement of program goals. This model is examined for its sensitivity to safe/unsafe failure rates, sybsystem redundant configurations, test and repair intervals, monitoring by reactor operators; and the control exercised over system reliability by design modifications and the selection of system operating characteristics. (U.S.)

  6. Safety and core design of large liquid-metal cooled fast breeder reactors

    Science.gov (United States)

    Qvist, Staffan Alexander

    In light of the scientific evidence for changes in the climate caused by greenhouse-gas emissions from human activities, the world is in ever more desperate need of new, inexhaustible, safe and clean primary energy sources. A viable solution to this problem is the widespread adoption of nuclear breeder reactor technology. Innovative breeder reactor concepts using liquid-metal coolants such as sodium or lead will be able to utilize the waste produced by the current light water reactor fuel cycle to power the entire world for several centuries to come. Breed & burn (B&B) type fast reactor cores can unlock the energy potential of readily available fertile material such as depleted uranium without the need for chemical reprocessing. Using B&B technology, nuclear waste generation, uranium mining needs and proliferation concerns can be greatly reduced, and after a transitional period, enrichment facilities may no longer be needed. In this dissertation, new passively operating safety systems for fast reactors cores are presented. New analysis and optimization methods for B&B core design have been developed, along with a comprehensive computer code that couples neutronics, thermal-hydraulics and structural mechanics and enables a completely automated and optimized fast reactor core design process. In addition, an experiment that expands the knowledge-base of corrosion issues of lead-based coolants in nuclear reactors was designed and built. The motivation behind the work presented in this thesis is to help facilitate the widespread adoption of safe and efficient fast reactor technology.

  7. Fusion breeder

    International Nuclear Information System (INIS)

    The fusion breeder is a fusion reactor designed with special blankets to maximize the transmutation by 14 MeV neutrons of uranium-238 to plutonium or thorium to uranium-233 for use as a fuel for fission reactors. Breeding fissile fuels has not been a goal of the US fusion energy program. This paper suggests it is time for a policy change to make the fusion breeder a goal of the US fusion program and the US nuclear energy program. The purpose of this paper is to suggest this policy change be made and tell why it should be made, and to outline specific research and development goals so that the fusion breeder will be developed in time to meet fissile fuel needs

  8. Gas core reactors for actinide transmutation and breeder applications. Annual report

    International Nuclear Information System (INIS)

    This work consists of design power plant studies for four types of reactor systems: uranium plasma core breeder, uranium plasma core actinide transmuter, UF6 breeder and UF6 actinide transmuter. The plasma core systems can be coupled to MHD generators to obtain high efficiency electrical power generation. A 1074 MWt UF6 breeder reactor was designed with a breeding ratio of 1.002 to guard against diversion of fuel. Using molten salt technology and a superheated steam cycle, an efficiency of 39.2% was obtained for the plant and the U233 inventory in the core and heat exchangers was limited to 105 Kg. It was found that the UF6 reactor can produce high fluxes (10 to the 14th power n/sq cm-sec) necessary for efficient burnup of actinide. However, the buildup of fissile isotopes posed severe heat transfer problems. Therefore, the flux in the actinide region must be decreased with time. Consequently, only beginning-of-life conditions were considered for the power plant design. A 577 MWt UF6 actinide transmutation reactor power plant was designed to operate with 39.3% efficiency and 102 Kg of U233 in the core and heat exchanger for beginning-of-life conditions

  9. Sodium technology for fast breeder reactors

    International Nuclear Information System (INIS)

    Sodium, because of its good heat transfer and nuclear properties, is used as a coolant in fast reactors. It is also used largely as a reducing agent in pharmaceutical, perfumery and general chemical industries. Its affinity to react with air and water is a strong disadvantage. However, this is fully understood and the design of engineering systems take care of this aspect. With several experimental and test facilities established over the years in this country as well as abroad, the 'sodium technology' has reached a level of maturity. The design of sodium systems considering all the physical and chemical properties and the developmental work carried out at Indira Gandhi Centre for Atomic Research are broadly covered in this report. (author)

  10. Advanced fuels for fast reactors

    International Nuclear Information System (INIS)

    Full text: In addition to traditional fast reactor fuels that contain Uranium and Plutonium, the advanced fast reactor fuels are likely to include the minor actinides [Neptunium (Np), Americium (Am) and Curium (Cm)]. Such fuels are also referred to as transmutation fuels. The goal of transmutation fuel development programs is to develop and qualify a nuclear fuel system that performs all of the functions of a traditional fast spectrum nuclear fuel while destroying recycled actinides. Oxide, metal, nitride, and carbide fuels are candidates under consideration for this application, based on historical knowledge of fast reactor fuel development and specific fuel tests currently being conducted in international transmutation fuel development programs. Early fast reactor developers originally favored metal alloy fuel due to its high density and potential for breeder operation. The focus of pressurized water reactor development on oxide fuel and the subsequent adoption by the commercial nuclear power industry, however, along with early issues with low burnup potential of metal fuel (now resolved), led later fast reactor development programs to favor oxide fuels. Carbide and nitride fuels have also been investigated but are at a much lower state of development than metal and oxide fuels, with limited large scale reactor irradiation experience. Experience with both metal and oxide fuels has established that either fuel type will meet performance and reliability goals for a plutonium fueled fast spectrum test reactor, both demonstrating burnup capability of up to 20 at.% under normal operating conditions, when clad with modified austenitic or ferritic martensitic stainless steel alloys. Both metal and oxide fuels have been shown to exhibit sufficient margin to failure under transient conditions for successful reactor operation. Summary of selected fuel material properties taken are provided in the paper. The main challenge for the development of transmutation fast reactor

  11. Contained fission explosion breeder reactor system

    International Nuclear Information System (INIS)

    A reactor system for producing useful thermal energy and valuable isotopes, such as plutonium-239, uranium-233, and/or tritium, in which a pair of sub-critical masses of fissile and fertile actinide slugs are propelled into an ellipsoidal pressure vessel. The propelled slugs intercept near the center of the chamber where the concurring slugs become a more than prompt configuration thereby producing a fission explosion. Re-useable accelerating mechanisms are provided external of the vessel for propelling the slugs at predetermined time intervals into the vessel. A working fluid of lean molten metal slurry is injected into the chamber prior to each explosion for the attenuation of the explosion's effects, for the protection of the chamber's walls, and for the absorbtion of thermal energy and debris from the explosion. The working fluid is injected into the chamber in a pattern so as not to interfere with the flight paths of the slugs and to maximize the concentration of working fluid near the chamber's center. The heated working fluid is drained from the vessel and is used to perform useful work. Most of the debris from the explosion is collected as precipitate and is used for the manufacture of new slugs

  12. Description of a materials/coolant laboratory for support of the Breeder Reactor Technology Shipping Program

    International Nuclear Information System (INIS)

    A description of a facility devoted to evaluating the environmental compatibility and mechanical response of materials suitable for a breeder reactor spent-fuel shipping cask is given. The facility presently consists of a closed-loop servo-controlled hydraulic, horizontal test system coupled to an environmental chamber, generalized mechanical test equipment and high-rate mechanical behavior apparatus. Future plans include the procurement of real-time computer control equipment which will be used to assess the effects of complex load-time histories on spent-fuel shipping cask materials

  13. Conceptual design of loop-in-tank type Indian molten salt breeder reactor concept

    International Nuclear Information System (INIS)

    The third stage of Indian nuclear power programme envisages use of thorium as fertile material with 233U, which is proposed to be obtained from reprocessing of spent fuel of Pu/Th based fast reactors in the later part of the second stage of the programme. In India, thorium based reactors have been designed in many configurations, from light water cooled designs to high temperature liquid metal and molten salt cooled options. Another option, which holds promise, is the molten salt-fuelled reactor, which can be configured to give significant breeding ratios. A crucial part for achieving reasonable breeding in such reactors is the need to reprocess the salt continuously, either online or in batch mode. India has recently started carrying out fundamental studies so as to arrive at a conceptual design of Indian Molten Salt Breeder Reactor (IMSBR). (author)

  14. IAMBUS, a computer code for the design and performance prediction of fast breeder fuel rods

    International Nuclear Information System (INIS)

    IAMBUS is a computer code for the thermal and mechanical design, in-pile performance prediction and post-irradiation analysis of fast breeder fuel rods. The code deals with steady, non-steady and transient operating conditions and enables to predict in-pile behavior of fuel rods in power reactors as well as in experimental rigs. Great effort went into the development of a realistic account of non-steady fuel rod operating conditions. The main emphasis is placed on characterizing the mechanical interaction taking place between the cladding tube and the fuel as a result of contact pressure and friction forces, with due consideration of axial and radial crack configuration within the fuel as well as the gradual transition at the elastic/plastic interface in respect to fuel behavior. IAMBUS can be readily adapted to various fuel and cladding materials. The specific models and material correlations of the reference version deal with the actual in-pile behavior and physical properties of the KNK II and SNR 300 related fuel rod design, confirmed by comparison of the fuel performance model with post-irradiation data. The comparison comprises steady, non-steady and transient irradiation experiments within the German/Belgian fuel rod irradiation program. The code is further validated by comparison of model predictions with post-irradiation data of standard fuel and breeder rods of Phenix and PFR as well as selected LWR fuel rods in non-steady operating conditions

  15. Operating and test experience with Experimental Breeder Reactor number 2 (EBR-II), the Integral Fast Reactor (IFR) prototype

    International Nuclear Information System (INIS)

    The Experimental Breeder Reactor number 2 (EBR-II) has operated for 30 years, the longest for any liquid metal cooled reactor (LMR) power plant in the world. Given the scope of what has been developed and demonstrated over those years, it is arguably the most successful test reactor operation ever. Tests have been carried out on virtually every fast reactor fuel type. The reactor itself has been extensively studied. The most dramatic safety tests, conducted on 3 April, 1986, showed that an LMR with metallic fuel could safely accommodate loss of flow or loss of heat-sink without scram. EBR-II operated as the Integral Fast Reactor (IFR) prototype, demonstrating important innovations in safety, plant design, fuel design and actinide recycle. The ability to accommodate anticipated transients without scram passively resulted in significant simplification of the reactor plant, primarily through less reliance on emergency power and not having to require the secondary sodium or steam systems to be safety grade. These features have been quantified in a probabilistic risk assessment (PRA) conducted for EBR-II, demonstrating considerable safety advantages over other reactor concepts. Fundamental to the superior safety and operating characteristics of this reactor is the metallic U-Pu-Zr alloy fuel. Performance of the fuel has been fully proven: achieved burnup levels exceed 20 at.% in the lead test assemblies. A complete set of fuel performance and safety limits has been developed and was carried forward in formal safety documents supporting conversion of the core to IFR fuel. The last major demonstration planned was to assess the performance of recycled actinides in the fuel and to confirm that passive safety characteristics are maintained with recycled actinide fuel in the core. (author)

  16. Fast breeder reactor. The past, the present and the future. (6) History of fast reactor development in Japan - 1

    International Nuclear Information System (INIS)

    History and present state of fast breeder reactor was reviewed in series. As a history of fast reactor development in Japan - 1, this sixth lecture presented the start of FBR development, and construction and operation of the experimental FBR (JOYO). The JOYO began operation in 1977 and now is being operated at 140 MWt after two times of upgraded modification. The JOYO is aimed at (1) advancement of technology through and experiment, (2) conducting irradiation tests on fuels and materials and (3) validation of innovative technology for development of a future FBR. (T. Tanaka)

  17. Summary of several hydraulic tests in support of the light water breeder reactor design (LWBR development program)

    International Nuclear Information System (INIS)

    As part of the Light Water Breeder Reactor development program, hydraulic tests of reactor components were performed. This report presents the results of several of those tests performed for components which are somewhat unique in their application to a pressurized water reactor design. The components tested include: triplate orifices used for flow distribution purposes, multiventuri type flowmeters, tight lattice triangular pitch rod support grids, fuel rod end support plates, and the balance piston which is a major component of the movable fuel balancing system. Test results include component pressure loss coefficients, flowmeter coefficients and fuel rod region pressure drop characteristics

  18. Experiences with fast breeder reactor education in laboratory and short course settings

    International Nuclear Information System (INIS)

    The breeder reactor industry throughout the world has grown impressively over the last two decades. Despite the uncertainties in some national programs, breeder reactor technology is well established on a global scale. Given the magnitude of this technological undertaking, there has been surprisingly little emphasis on general breeder reactor education - either at the university or laboratory level. Many universities assume the topic too specialized for including appropriate courses in their curriculum - thus leaving students entering the breeder reactor industry to learn almost exclusively from on-the-job experience. The evaluation of four course presentations utilizing visual aids is presented

  19. Computational intelligent systems for Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Nearly 15000 process signals are digitized by physically and functionally distributed embedded systems in Prototype Fast Breeder Reactor (PFBR). Digitized signals are processed and relevant information is displayed through Large video display systems at Control Room. It is necessary that correct and reliable information need to be provided to the plant operator. Computational intelligent systems play a major role in enhancing the safe operation of the Nuclear reactor. The paper explains the features of three such systems, one for on-line validation of neutronic power channel through on-line thermal balance calculation and another for detection of anomalous reactivity addition through on-line reactivity balance computation and third for on-line computation of Reactor power from fluctuations of core thermocouple signals. (author)

  20. Clinch River Breeder Reactor secondary control rod system

    International Nuclear Information System (INIS)

    The shutdown system for the Clinch River Breeder Reactor (CRBR) includes two independent systems--a primary and a secondary system. The Secondary Control Rod System (SCRS) is a new design which is being developed by General Electric to be independent from the primary system in order to improve overall shutdown reliability by eliminating potential common-mode failures. The paper describes the status of the SCRS design and fabrication and testing activities. Design verification testing on the component level is largely complete. These component tests are covered with emphasis on design impact results. A prototype unit has been manufactured and system level tests in sodium have been initiated

  1. Base isolation system for prototype fast breeder reactor

    International Nuclear Information System (INIS)

    The use of seismic isolation specially in the high seismic regions has gained increasing interest as a viable and efficient solution to earthquake ground motion both within and outside the nuclear field. A feasibility study to see the effect of laminated rubber bearing pads for the 500 MWe pool type fast breeder reactor has been carried out. The results show that there is 2 to 2.5 times reduction in floor response spectra peak and the seismic loads on the components are considerably reduced. The problem areas include the potential for the large sloshing amplitudes, accommodating large displacements in the piping etc. (author)

  2. Innovations in Equipment Erection of Prototype Fast Breeder Reactor (PFBR)

    International Nuclear Information System (INIS)

    Prototype Fast Breeder Reactor (PFBR) is sodium cooled, pool type reactor with generating capacity of 1250 MWt/500 MWe. Reactor assembly consists of large dimensional vessels like Safety vessel (13.54 m diameter, 12.8 m height and weight approximately 155 MT) and Main vessel (12.9 m diameter, 12.94 m height and weight approximately 202 MT including core catcher, core support structure and cooling pipes) and Steam generator (26 m length, 1.5 m diameter, and weight approximately 35 MT). PFBR reactor equipment erection was a challenging task where thin walled vessels had transported and handled with utmost precaution to avoid radial forces on the vessels which could buckle the vessels. There was a real challenge in lifting the vessels without swing, placement of large size and heavy vessel at a distance of 57 m where the crane operator had no line of site to the equipment being erected. To handle such over dimensional reactor components many mock-up tests had been carried out before erection and gained lot of confidence. Lot of care had been taken during lifting, handling and erection of thin walled over dimensional reactor components with innovative methods used for lifting fixtures, guiding arrangements, alignment fixtures and achieved the stringent erection tolerances. This paper discusses the first ever experiences gained during the handling and erection of such thin walled, over dimensional reactor components at PFBR site. (author)

  3. Future fuel cycle development for CANDU reactors

    International Nuclear Information System (INIS)

    The CANDU reactor has proven to be safe and economical and has demonstrated outstanding performance with natural uranium fuel. The use of on-power fuelling, coupled with excellent neutron economy, leads to a very flexible reactor system with can utilize a wide variety of fuels. The spectrum of fuel cycles ranges from natural uranium, through slightly enriched uranium, to plutonium and ultimately thorium fuels which offer many of the advantages of the fast breeder reactor system. CANDU can also burn the recycled uranium and/or the plutonium from fuel discharged from light water reactors. This synergistic relationship could obviate the need to re-enrich the reprocessed uranium and allow a simpler reprocessing scheme. Fule management strategies that will permit future fuel cycles to be used in existing CANDU reactors have been identified. Evolutionary design changes will lead to an even greater flexibility, which will guarantee the continued success of the CANDU system. (author)

  4. Hot Fuel Examination Facility's neutron radiography reactor

    International Nuclear Information System (INIS)

    Argonne National Laboratory-West is located near Idaho Falls, Idaho, and is operated by the University of Chicago for the United States Department of Energy in support of the Liquid Metal Fast Breeder Reactor Program, LMFBR. The Hot Fuel Examination Facility, HFEF, is one of several facilities located at the Argonne Site. HFEF comprises a large hot cell where both nondestructive and destructive examination of highly-irradiated reactor fuels are conducted in support of the LMFBR program. One of the nondestructive examination techniques utilized at HFEF is neutron radiography, which is provided by the NRAD reactor facility (a TRIGA type reactor) below the HFEF hot cell

  5. Conceptual design of Indian molten salt breeder reactor

    Indian Academy of Sciences (India)

    P K Vijayan; A Basak; I V Dulera; K K Vaze; S Basu; R K Sinha

    2015-09-01

    The third stage of Indian nuclear power programme envisages the use of thorium as the fertile material with 233U, which would be obtained from the operation of Pu/Th-based fast reactors in the later part of the second stage. Thorium-based reactors have been designed in many configurations, from light water-cooled designs to high-temperature liquid metal-cooled options. Another option, which holds promise, is the molten salt-fuelled reactor, which can be configured to give significant breeding ratios. A crucial part for achieving reasonable breeding in such reactors is the need to reprocess the salt continuously, either online or in batch mode. India has recently started carrying out fundamental studies so as to arrive at a conceptual design of Indian molten salt breeder reactor (IMSBR). Presently, various design options and possibilities are being studied from the point of view of reactor physics and thermal hydraulic design. In parallel, fundamental studies on natural circulation and corrosion behaviour of various molten salts have also been initiated.

  6. Fast breeder reactor core concept consistent with fuel cycle system during the transition period from LWR to FBR cycles in Japan

    International Nuclear Information System (INIS)

    During the transition period from the LWR cycle to a FBR cycle, the Pu needed for FBR start-up will be obtained from the next reprocessing facility. Therefore, from the economic viewpoint, it would be better if the FBR core in the next reprocessing facility had a lower Pu inventory. A design concept study was carried out for an FBR core with a lower Pu inventory that is consistent with the fuel cycle system during the transition period. The Pu from LWR spent fuel will be used for FBR start-up at least for the initial core cycle and succeeding first few cycles during the transition period. An FBR core loaded with Pu from LWR spent fuel has higher burnup reactivity than one loaded with Pu from FBR multi-recycling fuel composition. The increased burnup reactivity may reduce the cycle length of the FBR core. We chose loading of minor actinides (MAs) into the FBR MOX fuel as the countermeasure to the increased burnup reactivity from the viewpoint of utilizing nuclear characteristics of the MAs. The maximum MA content in the MOX fuel was set to 5% on the basis of irradiation test results. MAs recovered from the LWR spent fuel that provides Pu for FBR start-up are loaded into the initial loading fuel assemblies and exchanged fuel assemblies for several cycles until equilibriumis reached. The average MA content of the initially loaded fuel was assumed to be 3%, and that of the exchanged fuel was assumed to be 5%. The core performance including burnup characteristics and reactivity coefficient were also evaluated, and it was confirmed that the transient core from the initial loading until the equilibrium cycle for loaded Pu from LWR spent fuels could maintain performance neary equal to that of an FBR multi-recycling core. (author)

  7. Breeder reactors: a technique at the service of humanity

    International Nuclear Information System (INIS)

    A genuine energy policy is not conceived purely for a short term. It must on the contrary take into consideration many national and international facts in order to arrive at a balance which takes into account both the interests of the country where it is to be applied and the future interests of humanity. Growth and energy consumption make a pair. Considering the forecasts of future consumption, a rational utilization of the energy sources is a priority. The rational utilization of the energy potentialities of uranium takes a prominent place in this priority. In the fission energy of the atoms, the breeder reactors are the only types which can give their full meanings to the words economy, ecology, rationality etc. In calling for innovation, the breeder reactors are the prime movers for an advanced industry and a guarantee for the future penetration of electricity in many fields. They are thus important elements for the creation of employment. This paper also deals with questions of international cooperation, non-proliferation and the necessity for disarmament

  8. UF6 breeder reactor power plants for electric power generation

    International Nuclear Information System (INIS)

    The reactor concept analyzed is a 233UF6 core surrounded by a molten salt (Li7F, BeF2, ThF4) blanket. Nuclear survey calculations were carried out for both spherical and cylindrical geometries. A maximum breeding ratio of 1.22 was found. Thermodynamic cycle calculations were performed for a variety of Rankine cycles. Optimization of a Rankine cycle for a gas core breeder reactor employing an intermediate heat exchanger gave a maximum efficiency of 37 percent. A conceptual design is presented along with a system layout for a 1000 MW stationary power plant. The advantages of the GCBR are as follows: (1) high efficiency, (2) simplified on-line reprocessing, (3) inherent safety considerations, (4) high breeding ratio, (5) possibility of burning all or most of the long-lived nuclear waste actinides, and (6) possibility of extrapolating the technology to higher temperatures and MHD direct conversion

  9. UF6 breeder reactor power plants for electric power generation

    Science.gov (United States)

    Rust, J. H.; Clement, J. D.; Hohl, F.

    1976-01-01

    The reactor concept analyzed is a U-233F6 core surrounded by a molten salt (Li(7)F, BeF2, ThF4) blanket. Nuclear survey calculations were carried out for both spherical and cylindrical geometries. Thermodynamic cycle calculations were performed for a variety of Rankine cycles. A conceptual design is presented along with a system layout for a 1000 MW stationary power plant. Advantages of the gas core breeder reactor (GCBR) are as follows: (1) high efficiency; (2) simplified on-line reprocessing; (3) inherent safety considerations; (4) high breeding ratio; (5) possibility of burning all or most of the long-lived nuclear waste actinides; and (6) possibility of extrapolating the technology to higher temperatures and MHD direct conversion.

  10. Characterizing the tribological behaviour of fast breeder reactor materials

    International Nuclear Information System (INIS)

    The object of these tests is to define the behaviour of material couples working in conditions as representative as possible of reactor operation. For this purpose a certain number of test installations have been developed to simulate the most typical cases of friction encountered: plane to plane geometry, rotational bearings, guiding bearings. Endurance tests have also been carried out on ball bearings and ballscrews samples. As said before, the test conditions attempt to reproduce as faithfully as possible the environment of the materials used in fast breeder reactors, particularly in: - using purified liquid sodium, and maintaining it isotherm, respectively at three temperature levels: 180, 400 and 5500C; - or using argon containing sodium aerosol particles. Some typical values of friction coefficients and rates of wear obtained during the tests with certain couples of materials are given here as examples. The aims which are currently guiding the direction of the tests are also briefly described

  11. Operating experience of fast breeder reactors in the USSR

    International Nuclear Information System (INIS)

    The operating experience results of BN-600, BN-350, BOR-60 and BR-10 fast breeder reactors are presented. The fast reactors design and operation experience in the USSR has demonstrated their high operational qualities, safety, reserves of improvement. After 11 years' operation the BN-600 and 18 years' operation the BN-350 these two nuclear plants present a very satisfactory global loading rate of above 65%. The operation flexibility of the nuclear power plants and, in particular, the possibility of operation at 2/3 nominal power (BN-600) and at 4/5 and/or 3/5 nominal power (BN-350) have allowed for these loading rates to be reached in spite of numerous steam generators and pumps replacement. (J.P.N.)

  12. A linear model of the Fast Breeder Test Reactor Plant

    International Nuclear Information System (INIS)

    A linear analysis of the Fast Breeder Test Reactor System, consisting of the reactor, intermediate heat exchanger, steam generator and connected piping is presented. The problem of variable boundaries in the steam generator is reduced to a problem of fixed boundaries by dividing the steam generator into six zones. Based upon this, one can obtain the transfer function of any input/output combination. Starting with the time domain non-linear partial differential equations, the problem is reduced to a system of linear equations in complex variables, which can be solved basically by Gaussian elimination process. The results of this work will be useful in determining a suitable control scheme for waterflow in the steam generator and the control parameters. (auth.)

  13. Optimisation of safety parameters in fast breeder test reactor

    International Nuclear Information System (INIS)

    Full text: Optimisation of safety parameters is an important aspect to be considered in the design of nuclear power plant and also becomes extremely important activity to be followed up during the commissioning and operating phases of the plant taking into account the operational feed back and review of incidental situations and available diversity and reliability. Otherwise, the spurious/ superfluous trips on the reactor besides affecting the availability of the plant, initiate plant transients causing stress for the plant equipment resulting in reduction of plant life. This activity has a significant role to play in attaining the maximum availability of the plant, without compromising safety. The study and evolution of optimisation process in fast breeder test reactor (FBTR); at Kalpakkam has been an interesting and rewarding experience

  14. Three core concepts for producing uranium-233 in commercial pressurized light water reactors for possible use in water-cooled breeder reactors

    International Nuclear Information System (INIS)

    Selected prebreeder core concepts are described which could be backfit into a reference light water reactor similar to current commercial reactors, and produce uranium-233 for use in water-cooled breeder reactors. The prebreeder concepts were selected on the basis of minimizing fuel system development and reactor changes required to permit a backfit. The fuel assemblies for the prebreeder core concepts discussed would occupy the same space envelope as those in the reference core but contain a 19 by 19 array of fuel rods instead of the reference 17 by 17 array. An instrument well and 28 guide tubes for control rods have been allocated to each prebreeder fuel assembly in a pattern similar to that for the reference fuel assemblies. Backfit of these prebreeder concepts into the reference reactor would require changes only to the upper core support structure while providing flexibility for alternatives in the type of fuel used

  15. A contribution to the analysis of the thermal behaviour of Fast Breeder fuel rods with UO2-PuO2 fuel

    International Nuclear Information System (INIS)

    The fuel of Fast Breeder Reactors which consists of Uranium and Plutonium dioxide is mainly characterized by the amount and distribution of void volume and Plutonium and the amount of oxygen. Irradiation experiments carried out with this fuel have shown that initial structure of the fuel pellet is subjected to large changes during operation. These are consequences of the radial and axial temperature gradients within the fuel rods. (Author) 54 refs

  16. Fast breeder reactor-block antiseismic design and verification

    International Nuclear Information System (INIS)

    The Specialists' Meeting on ''Fast Breeder Reactor-Block Antiseismic Design and Verification'' was organized by the ENEA Fast Reactor Department in co-operation with the International Working Group (IWGFR) of the International Atomic Energy Agency (IAEA), according to the recommendations of the 19th IAEA/IWGFR Meeting. It was held in Bologna, at the Headquarters of the ENEA Fast Reactor Department, on October 12-15, 1987, in the framework of the Celebrations for the Ninth Centenary of the Bologna University. The proceedings of the meeting consists of three parts. Part 1 contains the introduction and general comments, the agenda of the meeting, session summaries, conclusions and recommendations and the list of participants. Part 2 contains 8 status reports of Member States participating in the Working Group. Contributed papers were published in Part 3 and were further subdivided into 5 sessions as follows: whole reactor-block analysis (4 papers); whole reactor-block analysis (sloshing and buckling, seismic isolation effects) (8 papers); detailed core analysis (6 papers); shutdown systems and core structural and functional verifications (6 papers); component and piping analysis (7 papers). A separate abstract was prepared for each of the 8 status reports and 31 contributed papers. Refs, figs and tabs

  17. It is now time to proceed with a gas-cooled breeder reactor (GBR) demonstration plant

    International Nuclear Information System (INIS)

    Since 1969, the GBRA has been making studies to provide evidence on questions which were not clear regarding the Gas-cooled Breeder Reactor: design feasibility and performance, safety, strategy and economics, and R and D necessary for a demonstration plant. Studies were carried out on a 1200-MW(e) commercial reference design with pin fuel, which was also used as a basis for a definition of the GBR demonstration plant. During the six years, a great deal of information has been generated at GBRA and it confirms the forecasts of the promoters of the Gas-cooled Breeder Reactor that the GBR is an excellent reactor from all points of view: design - the reactor can be engineered without major difficulty, using present techniques. As far as fuel is concerned, LMFBR fuel technology is applicable plus limited specific development effort. Performance - the GBR is the best breeder with oxide fuel and using conventional techniques. The strategy studies carried out at GBRA have clearly shown that a high performance breeder such as the GBR is absolutely required in large quantities by the turn of the century in order to avoid dependence on natural uranium resources. Regarding safety, a major step forward has been made when an ad hoc group on GBR safety, sponsored by the EEC, concluded that no major difficulties were anticipated which would prevent the GBR reaching adequate safety standards. Detailed economic assessments performed on an item-to-item basis have shown that the cost of a GBR with its high safety standard is about the same as that of an HTR. One can therefore conclude that, with the present cost of natural uranium, the GBR is competitive with the LWRs. Owing to the very limited R and D effort necessary and the obvious safety, economic and strategic advantages of the concept, it is concluded that the development and construction of a GBR demonstration plant must be started now if one wants to secure an adequate energy supply past the turn of the century. (author)

  18. Preliminary design of a Binary Breeder Reactor; Diseno preliminar de un reactor esferico de quema/cria

    Energy Technology Data Exchange (ETDEWEB)

    Garcia C, E. Y.; Francois, J. L.; Lopez S, R. C., E-mail: eliasgarcerv@hotmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac No. 8532, 62550 Jiutepec, Morelos (Mexico)

    2014-10-15

    A binary breeder reactor (BBR) is a reactor that by means of the transmutation and fission process can operates through the depleted uranium burning with a small quantity of fissile material. The advantages of a BBR with relation to other nuclear reactor types are numerous, taking into account their capacity to operate for a long time without requiring fuel reload or re-arrangement. In this work four different simulations are shown carried out with the MCNPX code with libraries Jeff-3.1 to 1200 K. The objective of this study is to compare two different models of BBR: a spherical reactor and a cylindrical one, using two fuel cycles for each one of them (U-Pu and Th-U) and different reflectors for the two different geometries. For all the models a super-criticality state was obtained at least 10.9 years without carrying out some fuel re-arrangement or reload. The plutonium-239 production was achieved in the models where natural uranium was used in the breeding area, while the production of uranium-233 was observed in the cases where thorium was used in the fertile area. Finally, a behavior of stationary wave reactor was observed inside the models of spherical reactor when contemplating the power uniform increment in the breeding area, while inside the cylindrical models was observed the behavior of a traveling wave reactor when registering the displacement of the burnt wave along the cylindrical model. (Author)

  19. Programme and status of the development of the fast breeder reactor system in France

    International Nuclear Information System (INIS)

    The fundamental justification for fast breeder reactors is based on their breeding ability, and the development of this reactor line is a matter of national importance because there is no reasonable and adequate alternative that could both cover the increasing future energy requirements and provide energy independence for France. The 250MW(e) Phenix power plant has been satisfactorily operated as a demonstration plant for two years now. This proof of the validity of the system was necessary. It underlined its reliability and confirmed the value of the process to the constructors and to the domestic utility, EDF, who confidently decided on industrial and commercial commitment to the new reactor line. This attitude is demonstrated in the following actions in France: (a) the construction by Nersa of the Super-Phenix power plant, a 1200MW(e) prototype commercial power plant which will provide experience in building and operating large fast-neutron power plants; (b) the firm intention of EDF to proceed and bring breeders into operation in the 1990s, representing about 8000MW(e); (c) the creation of a strong industrial organization: the creation in France of the Novatome Company corresponds to this objective - Novatome is now responsible for the industrial development of fast breeder reactors with Creusot-Loire as industrial basis and with all the know-how owned by CEA; (d) the Compagnie generale des matieres nucleaires, Cogema, a CEA subsidiary, is reviewing the actions necessary to reach industrial levels for the cost of the whole fuel cycle. On the other hand, the negotiation of international agreements adds the knowledge and experience gained in the R and D programmes of the various countries concerned. They improve the value and the technical, industrial and commercial basis of this new type of reactor, thus enlarging the prospects of development and commercialization. (author)

  20. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    An improved nuclear power reactor fuel element is described which consists of fuel rods, rod guide tubes and an end plate. The system allows direct access to an end of each fuel rod for inspection purposes. (U.K.)

  1. Economic performance of liquid-metal fast breeder reactor and gas-cooled fast reactor radial blankets

    International Nuclear Information System (INIS)

    The economic performance of the radial blanket of a liquid-metal fast breeder reactor (LMFBR) and a gas-cooled fast reactor (GCFR) has been studied based on the calculation of the net financial gain as well as the value of the levelized fuel cost. The necessary reactor physics calculations have been performed using the code CITATION, and the economic analysis has been carried out with the code ECOBLAN, which has been written for that purpose. The residence time of fuel in the blanket is the main variable of the economic analysis. Other parameters that affect the results and that have been considered are the value of plutonium, the price of heat, the effective cost of money, and the holdup time of the spent fuel before reprocessing. The results show that the radial blanket of both reactors is a producer of net positive income for a broad range of values of the parameters mentioned above. The position of the fuel in the blanket and the fuel management scheme applied affect the monetary gain. There is no significant difference between the economic performance of the blanket of an LMFBR and a GCFR

  2. Design of a redundant-load-path lifting fixture for the Clinch River Breeder Reactor Plant spent-fuel shipping cask

    International Nuclear Information System (INIS)

    A newly developed concept for a redundant-load-path lifting fixture for spent fuel shipping casks is presented. The concept addresses remote attachment of the crane hook to the lifting fixture and remote attachment of the lifting fixture to this cask, and also allows the cask to be rotated from its shipping position to the vertical and lifted from its transport vehicle without requiring hands-on action

  3. Sodium and steam generator leak detection for prototype fast breeder reactor (PFBR)

    International Nuclear Information System (INIS)

    The construction of the Prototype Fast Breeder Reactor (PFBR) a 500 MWe pool type sodium cooled breeder reactor with MOX fuel has started at Kalpakkam. The Instrumentation and Control of PFBR is designed for safe, reliable and economic operation of the plant. Special feature of breeder reactors is sodium instrumentation. Leaks in sodium systems have the possibility of being exceptionally hazardous due to the reaction of liquid sodium with oxygen and water vapour in the air. In addition, leakage from primary systems can cause radioactive contamination. Potential regions of leakage are near welds and high stress areas. Sodium also reacts with concrete releasing hydrogen and leading to damage and loss of strength of concrete structures. Leaking sodium catches fire depending on its temperature. Sodium temperature in the plant ranges from 423 K at filling condition to 820 K at reactor nominal power operating condition. Leak detectors are provided on pipelines, tanks and other capacities. Sodium leak detection systems are designed to meet requirements of ASME section XI- division 3 which specifies that sodium leak at the rate of 100 g/h are to be detected in 20 h for air filled vaults and 250 h for inert vaults. Diverse leak detection methods are employed for active and non-active sodium equipment and pipes. For detection of water leaks into Sodium in steam generators, Hydrogen in Sodium Detectors (HSD) are used. Hydrogen in Argon Detectors (HAD) are used for sodium temperatures below 623 K as HSD is not effective below this temperature due to non-dissolution of hydrogen formed. Choice and challenges posed in implementation of above leak detection requirements are discussed in this paper. (authors)

  4. Conceptual design of a passively safe thorium breeder Pebble Bed Reactor

    International Nuclear Information System (INIS)

    Highlights: • This work proposes three possible designs for a thorium Pebble Bed Reactor. • A high-conversion PBR (CR > 0.96), passively safe and within practical constraints. • A thorium breeder PBR (220 cm core) in practical regime, but not passively safe. • A passively safe breeder, requiring higher fuel reprocessing and recycling rates. - Abstract: More sustainable nuclear power generation might be achieved by combining the passive safety and high temperature applications of the Pebble Bed Reactor (PBR) design with the resource availability and favourable waste characteristics of the thorium fuel cycle. It has already been known that breeding can be achieved with the thorium fuel cycle inside a Pebble Bed Reactor if reprocessing is performed. This is also demonstrated in this work for a cylindrical core with a central driver zone, with 3 g heavy metal pebbles for enhanced fission, surrounded by a breeder zone containing 30 g thorium pebbles, for enhanced conversion. The main question of the present work is whether it is also possible to combine passive safety and breeding, within a practical operating regime, inside a thorium Pebble Bed Reactor. Therefore, the influence of several fuel design, core design and operational parameters upon the conversion ratio and passive safety is evaluated. A Depressurized Loss of Forced Cooling (DLOFC) is considered the worst safety scenario that can occur within a PBR. So, the response to a DLOFC with and without scram is evaluated for several breeder PBR designs using a coupled DALTON/THERMIX code scheme. With scram it is purely a heat transfer problem (THERMIX) demonstrating the decay heat removal capability of the design. In case control rods cannot be inserted, the temperature feedback of the core should also be able to counterbalance the reactivity insertion by the decaying xenon without fuel temperatures exceeding 1600 °C. Results show that high conversion ratios (CR > 0.96) and passive safety can be combined in

  5. Designing a SCADA system simulator for fast breeder reactor

    Science.gov (United States)

    Nugraha, E.; Abdullah, A. G.; Hakim, D. L.

    2016-04-01

    SCADA (Supervisory Control and Data Acquisition) system simulator is a Human Machine Interface-based software that is able to visualize the process of a plant. This study describes the results of the process of designing a SCADA system simulator that aims to facilitate the operator in monitoring, controlling, handling the alarm, accessing historical data and historical trend in Nuclear Power Plant (NPP) type Fast Breeder Reactor (FBR). This research used simulation to simulate NPP type FBR Kalpakkam in India. This simulator was developed using Wonderware Intouch software 10 and is equipped with main menu, plant overview, area graphics, control display, set point display, alarm system, real-time trending, historical trending and security system. This simulator can properly simulate the principle of energy flow and energy conversion process on NPP type FBR. This SCADA system simulator can be used as training media for NPP type FBR prospective operators.

  6. Status of fast breeder reactor development in Germany

    International Nuclear Information System (INIS)

    The KNK, the sodium cooled compact reactor is an experimental nuclear power plant of 20 MW electric power. Since 1977, it has been operated with fast reactor cores as KNK II. The KNK II/3 core was designed. The core fabrication has been largely completed. In 1990, the KNK II plant achieved a time availability of 56%. On January 8, 1991 KNK II was shut down for inspection. Since pre-nuclear commissioning was completed the Kalkar Nuclear Power Station SNR 300 has been operated in a mode similar to that of a power station. In March 1991 the financing partners decided not to prolong the standby phase because they do not think that the last construction permit and the operation permit will be issued within a definite period of time. The partners were convinced that the lack of progress in the licensing procedure was not caused by basic safety deficiencies of the project but by the way the licensing procedure was executed. The German fast breeder programme is now concentrated on contributions to the European Fast Reactor. (author)

  7. Developments in reprocessing technology for high-temperature and fast-breeder fuels

    International Nuclear Information System (INIS)

    The paper reports on the comprehensive long-range programmes in the USA to develop fuel-cycle technology for high-temperature gas-cooled reactors (HTGRs) and liquid-metal fast-breeder reactors (LMFBRs) which are being planned and implemented. In each case, the fuel cycle is an essential integral part of the power-reactor system, and the technical problems in the fuel cycle are being solved in parallel with the reactor development programme. For the LMFBR, the end point of the technical development will be a demonstration in a hot pilot plant to be operated with fully irradiated fuels from the Clinch River Breeder Reactor (CRBR) in the late 1980s. The programme is at present addressing process R and D problems, equipment development, and conceptual design of the pilot plant. The basic process will use a modified chop-leach head-end followed by Purex solvent-extraction cycles. Although a large base of experience will come from the early LWR plants, the significant differences between LMFBR and LWR fuels - sodium contamination, more complex fuel hardware, higher fissile and fission product content - introduce new problems requiring substantial development work. Increasing public concern and more stringent emission limits have prompted considerable effort toward retention of the gaseous fission products. For the HTGR, development is directed primarily toward the 233U-Th fuel cycle with a final goal of operation of a HTGR Fuel Recycle Demonstration Facility that can support the recycle requirements of 20-GW(e) installed capacity. The reference fuel element is a prismatic graphite block containing coated fissile and fertile particles bound together by a carbon matrix. Start-up is with 235U, which is replaced on subsequent cycles with the bred 233U. Spent fuel elements are crushed and burned, and the fuel particle oxides are dissolved and processed by solvent extraction. Refabrication must be done remotely because of γ-activity from 232U daughters, and the programme

  8. Physics calculations for the Clinch River Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kalimullah; Kier, P.H.; Hummel, H.H.

    1977-06-01

    Calculations of distributions of power and sodium void reactivity, unvoided and voided Doppler coefficients and steel and fuel worths have been performed using diffusion theory and first-order perturbation theory for the LWR discharge Pu-fueled CRBR at BOL, the FFTF-grade Pu-fueled CRBR at BOL and for the beginning and end of equilibrium cycle of the LWR-Pu-fueled CRBR. The results of the burnup and breeding ratio calculations performed for obtaining the reactor compositions during the equilibrium cycle are also reported. Effects of sodium and steel contents on the distributions of sodium void reactivity and steel worth have also been studied. Errors and uncertainties in the reactivity coefficients due to cross-sections and the two-dimensional geometric representations of the reactor used in the calculations have also been estimated. Comparisons of the results with those in the CRBR PSAR are also discussed.

  9. Fast-core thermal-blanket breeder reactor

    International Nuclear Information System (INIS)

    A preliminary assessment of the performance expected from a specific type of FCTB reactor, consisting of a gas-cooled fast system for the core and natural-uranium light-water thermal system for the blanket is reported. Both the core and the blanket use the 238U-Pu fuel cycle. When all the neutrons leaking out of the core reach the blanket, the blanket-to-core power ratio is estimated to be about 1.3. By reducing its water-to-fuel volume ratio below 1.5, the light water blanket can be designed to have a higher ksub(eff), while maintaining an equilibrium fissile fuel content. Compared with conventional FBRs, having the same power output, the FCTB reactor considered offers the following advantages: a lower fissile fuel content, easier and safer control, no need for Pu separation. (B.G.)

  10. Anticipated transients without scram for light water reactors: implications for liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    In the design of light water reactors (LWRs), protection against anticipated transients (e.g., loss of normal electric power and control rod withdrawal) is provided by a highly reliable scram, or shutdown system. If this system should become inoperable, however, the transient could lead to a core meltdown. The Nuclar Regulatory Commission (NRC) has proposed, in NUREG-0460 [1], new requirements (or acceptance criteria) for anticipated transients without scram (ATWS) events and the manner in which they could be considered in the design and safety evaluation of LWRs. This note assesses the potential impact of the proposed LWR-ATWS criteria on the liquid metal fast breeder reactor (LMFBR) safety program as represented by the Clinch River Breeder Reactor Plant

  11. Operating safety experience of fast breeder test reactor

    International Nuclear Information System (INIS)

    Full text: Operational safety criteria for nuclear reactors are very stringent and it is essential to incorporate adequate inherent and engineered safety features in the design to ensure safe operation of the reactor. Commissioning and operation of FBTR, being first of its kind in India based on nuclear chain reaction maintained by fast neutrons and use of high temperature liquid sodium as coolant, was a challenging task. Safe operation of the reactor for the past 17 years with good performance of sodium systems and the indigenous plutonium rich carbide fuel, touching a burn up level of 100 GWd/t has underlined the high level of design and operation competence achieved

  12. Fast breeder reactor blanket management: comparison of LMFBR and GCFR blankets

    International Nuclear Information System (INIS)

    The economic performance of the fast breeder reactor blanket, considering different fuel management schemes was studied. To perform this, the investigation started with a standard reactor physics calculation. Then, two economic models for evaluation of the economic performance of the radial blanket were developed. These models formed the basis of a computer code, ECOBLAN, which computes the net economic gain and the levelized fuel cost due to the radial blanket. The net gain in terms of dollars and $/kgHM-y and the levelized fuel cost in mills/kWhe were obtained as a function of blanket thickness and a residence time of the fuel in the blanket. A LMFBR and a GCFR were the reactor models considered in this study. The optimum radial blanket of a GCFR consists of two rows, that of a LMFBR consists of three rows. Regarding the different fuel management schemes, the fixed blanket was found to be more favorable than reshuffled blanket. Out-in and in-out reshuffled blanket offer almost the same net gain. In all the cases, the burnup calculated for the fuel was found to be less than the acceptable limit. There is an optimum residence time for the fuel in the blanket which depends on the position of the fuel in the blanket and the fuel management scheme studied. As expected, except for very short residence times (less than 2.5 years), the radial blanket is a net income producer. There is no significant difference between the economic performance of the blanket of a LMFBR and a GCFR

  13. High-temperature and breeder reactors - economic nuclear reactors of the future

    International Nuclear Information System (INIS)

    The thesis begins with a review of the theory of nuclear fission and sections on the basic technology of nuclear reactors and the development of the first generation of gas-cooled reactors applied to electricity generation. It then deals in some detail with currently available and suggested types of high temperature reactor and with some related subsidiary issues such as the coupling of different reactor systems and various schemes for combining nuclear reactors with chemical processes (hydrogenation, hydrogen production, etc.), going on to discuss breeder reactors and their application. Further sections deal with questions of cost, comparison of nuclear with coal- and oil-fired stations, system analysis of reactor systems and the effect of nuclear generation on electricity supply. (C.J.O.G.)

  14. Fuel sub-assemblies for nuclear reactors

    International Nuclear Information System (INIS)

    A fuel assembly for a liquid metal cooled fast breeder nuclear reactor comprises a bundle of spaced fuel pins within a tubular wrapper or sleeve. The wrapper is extended at one end by a tubular neutron shield of massive steel and the other end, has a spike extension whereby the sub-assembly can be located by plugging into a support structure. The invention provides that lateral displacement of individual fuel pin-containing wrappers to accommodate dimensional changes within the fuel assembly is effected by movement of each wrapper relative to its spike extension. (author)

  15. Design of Central Sub Assembly Temperature Monitoring System for Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    The Central Sub Assembly Temperature Monitoring (CSATM) System for 500 MWe Prototype Fast Breeder Reactor (PFBR) is a safety critical system. It is an independent, standalone, hardwired and diversified system to neutronic parameters. The detection of integrity of the subassembly plays a major role, because of high power density and compact core structure of PFBR fuel. To achieve this, CSATM system is provided for the measurement and detection of overshoot for Central Sub Assembly temperature. It protects the reactor from various incidents such as transient overpower at low power and high power, blockage of coolant, pipe rupture etc. CSATM system with triple modular redundancy is employed to measure the central sub-assembly outlet temperature (θCSA) and safety action will be initiated if temperature reaches beyond SCRAM threshold level. (author)

  16. Fast current pulse amplifier for neutron flux monitoring system of Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    The neutron flux monitoring system (NFMS) for Prototype Fast Breeder Reactor (PFBR) measures the neutron power and the reactivity changes in the core in all the states such as shut down, fuel handling, reactor startup, intermediate and power ranges using high temperature cylindrical fission chambers, four section fission counter and high temperature boron coated counter. Fast Current Pulse Amplifier has been developed to use in NFMS of PFBR that amplifies single/four numbers of input current pulses independently, discriminates and electronically wire - OR them to give differential pulse output along with the Campbell output. The paper describes the design, development of integrated single/Quad channel fast current pulse amplifier based on in-house developed ASIC, Hybrid IC, in built test features, LV and HV supplies. (author)

  17. Flow induced vibrations in liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    Flow induced vibrations are well known phenomena in industry. Engineers have to estimate their destructive effects on structures. In the nuclear industry, flow induced vibrations are assessed early in the design process, and the results are incorporated in the design procedures. In many cases, model testing is used to supplement the design process to ensure that detrimental behaviour due to flow induced vibrations will not occur in the component in question. While these procedures attempt to minimize the probability of adverse performance of the various components, there is a problem in the extrapolation of analytical design techniques and/or model testing to actual plant operation. Therefore, sodium tests or vibrational measurements of components in the reactor system are used to provide additional assurance. This report is a general survey of experimental and calculational methods in this area of structural mechanics. The report is addressed to specialists and institutions in industrialized and developing countries who are responsible for the design and operation of liquid metal fast breeder reactors. 92 refs, 90 figs, 8 tabs

  18. Present status and problems of development of fast breeder reactors

    International Nuclear Information System (INIS)

    The development of FBRs in Japan now reached the stage to conclude on the development organization for a demonstration reactor positioning one step before a practical reactor. FBRs can be operated while converting uranium-238 existing in natural uranium by 99.3% to fissile plutonium-239, as the result, the nuclear fuel more than that consumed can be produced. However, there are various technical difficulties in FBRs, and the construction cost is estimated to be considerably higher as compared with that of LWRs. Also the plutonium obtained by reprocessing spent fuel is used for FBRs, accordingly, the development of FBRs is inseparable from the establishment of nuclear fuel cycle. In order to get rid of the burden of enormous development cost for FBRs, the trend of international joint development is conspicuous. The Superphenix with 1200 MWe output under construction centering around SERENA is expected to attain the criticality in the spring of 1985. For the development of a demonstration reactor, it is necessary to increase the role of private businesses, and the smooth transfer of know-how accumulated in Power Reactor and Nuclear Fuel Development Corp. to civilian side is an important problem. (Kako, I.)

  19. Network representation of design knowledge of prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    A method of design knowledge representation was studied for the Japanese fast breeder reactor Monju, aiming at enhanced understanding of engineering considerations with mutual relations. Taking over design knowledge of Monju to next generation designers/engineers to be in charge of design of future FRs is by no means easy, in contrast with operation and maintenance knowledge which can be acquired in the real plant operation and maintenance. Specifications of the as-is Monju contains only a small part of the entire design knowledge, mainly by two reasons. Firstly, reasons for selecting the as-is specifications can not be understood until reaching proper knowledge source. Secondly, there are many passed-over options on the design specifications. Reasons for passing-over these options are not always technical inferiority. A large part of the current specifications are selected because the worst possible technical value can be foreseeable or guaranteed to be acceptable within limited R and D period and resource, not because the expected value is estimated to be the lower. In other words, in the future where new materials with improved properties, faster and more accurate analysis/prediction methods, rationalized technical standards or regulatory requirements, and/or some other environment for thorough comparison among specification options are available, these passed-over options are likely to be worth reconsidering. There are a huge number of technical documents on diversified engineering studies, such as calculation of maximum possible temperature gradient of important structures, necessary sodium flow rate in particular sub-assemblies, etc. for validation of each decision making in design. A large part of these documents are scanned and stored in a data base with each catalogue data for electronic browse. The authors propose a network representation of these items of design decision making, where the items are mutually connected by directed arcs, where nodes stand

  20. Fuel rod for nuclear reactors

    International Nuclear Information System (INIS)

    The fuel or breeder element with fission gas plenum has hollow spaces on the central part of the fuel or breeder material volume, which are filled with sodium. During operation, this sodium provides a second heat transport mechanism, as the sodium evaporates, rises into the fission gas plenum, condenses on the metal sleeve there and returns to the fuel or breeding zone under gravity or capillary effect. (DG)

  1. Current status of experimental breeder reactor-II [EBR-II] shutdown planning

    International Nuclear Information System (INIS)

    The Experimental Breeder Reactor--II (EBR-II) at Argonne National Laboratory--West (ANL-W) in Idaho, was shutdown in September, 1994 as mandated by the US Department of Energy. This sodium cooled reactor had been in service since 1964, and was to be placed in an industrially and radiologically safe condition for ultimate decommissioning. The deactivation of a liquid metal reactor presents unique concerns. The first major task associated with the project was the removal of all fueled assemblies. In addition, sodium must be drained from systems and processed for ultimate disposal. Residual quantities of sodium remaining in systems must be deactivated or inerted to preclude future hazards associated with pyrophoricity and generation of potentially explosive hydrogen gas. A Sodium Process Facility was designed and constructed to react the elemental sodium from the EBR-II primary and secondary systems to sodium hydroxide for disposal. This facility has a design capacity to allow the reaction of the complete inventory of sodium at ANL-W in less than two years. Additional quantities of sodium from the Fermi-1 reactor are also being treated at the Sodium Process Facility. The sodium environment and the EBR-II configuration, combined with the radiation and contamination associated with thirty years of reactor operation, posed problems specific to liquid metal reactor deactivation. The methods being developed and implemented at EBR-II can be applied to other similar situations in the US and abroad

  2. In-reactor experiments in fast breeder test reactor for assessment of core structural materials

    International Nuclear Information System (INIS)

    Fast Breeder Test Reactor (FBTR) at Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, India is a sodium cooled reactor with neutron flux level of the order of 1015 n/cm2/s and temperature of coolant in the range of 650-790K (380-520oC). This reactor is being used as a test bed for the development of fuel and structural materials required for Indian Fast Reactor Programme. FBTR is also used as a test facility to carry out accelerated irradiation tests on thermal reactor structural materials. In-reactor experiments on core structural materials are being carried out by subjecting prefabricated specimens to desired conditions of temperature and neutron fluence levels in FBTR. Non-instrumented irradiation capsules that can be loaded at any location of FBTR core are used for the experiments. Pressurised capsules of zirconium alloys have been developed and subjected to irradiation in FBTR to determine the irradiation creep rate of indigenously developed zirconium alloys (Zircaloy-2 and Zr-2.5%Nb alloy) for life assessment of pressure tubes of Indian Pressurised Heavy Water Reactors (PHWRs). Technology development of pressurised capsules was carried out at IGCAR. These pressurised capsules were filled with argon and a small fraction of helium at a high pressure (5.0-6.5 MPa at room temperature) in such a way that the target stresses were attained in the walls of the pressurised capsules at the desired temperature of irradiation in the reactor. FBTR was operated at a low power of 8 MWt during this irradiation campaign to have an inlet temperature of about 579 K (306oC) which was close to the temperature of pressure tubes at full power in PHWR. Irradiation of thirty pressurised capsules was carried out in FBTR using six irradiation capsules for different durations (upto 79 days). The fluence levels attained by the pressurised capsules were up to 1.1 x 1021 n/cm2 (E> 1 MeV) at temperatures of 579 to 592 K. Post-irradiation increase in diameter of the pressurised

  3. Development of standards and investigation of safety examination items for advancement of safety regulation of fast breeder reactor

    International Nuclear Information System (INIS)

    The purposes of this study are to prepare the fuel technical standard and the structure and materials standard of fast breeder reactors (FBRs), and to develop the requirements in a reactor establishment permission. The objects of this study are mainly the Monju high performance core and a demonstration FBR. In JFY 2012, the following results were obtained. As for the fuel technical standard, the fuel technical standard adapting the examination of integrity of the FBR fuels was prepared based on the information and data obtained in this study. As for the structure and material standard, the investigation of the revised parts of the standard was carried out. And as for the examination of the safety requirements, safety evaluation items for the future FBR plant and the fission products to be considered in a reactor establishment permission were investigated and examined. (author)

  4. Reactor shutdown system of prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Full text: The shutdown system of PFBR is designed to assure a very high reliability by employing well known principles of redundancy, diversity and independence. The failure probability of the shutdown system limited to -6/ ry. Salient features of the shutdown system are: Two independent shutdown systems, each of them able to accommodate an additional single failure and made up of a trip system and an associated absorber rod group. Diversity between trip systems, rods and mechanisms. Initiation of SCRAM by two diverse physical parameters of the two shutdown systems for design events leading potentially to unacceptable conditions is the core. The first group of nine rods called control and safety rods (CSR) is used for both shutdown as well as power regulation. The second group consisting of three rods known as diverse safety rods (DSR) is used only for shutdown. Diversity between the two groups is ensured by varying the operating conditions of the electromagnets and the configurations of the mobile parts. The reactivity worth of the absorber rods have been chosen such that each group of rods would ensure cold shutdown on SCRAM even when the most reactive rod of the group fails to drop. Together the two groups ensure a shutdown margin of 5000 pcm. The speed and individual rod worth of the CSR is chosen from operational and safety considerations during reactor start up and raising of power. Required drop time of rods during SCRAM depends on the incident considered. For a severe reactivity incident of 3 $/s this has to be limited to 1s and is ensured by limiting electromagnet response time and facilitating drop by gravity. Design safety limits for core components have been determined and SCRAM parameters have been identified by plant dynamic analysis to restrict the temperatures of core components within the limits. The SCRAM parameters are distributed between the two systems appropriately. Fault tree analysis of the system has been carried out to determine the

  5. An Evaluation of liquid metal leak detection methods for the Clinch River Breeder Reactor Plant

    Energy Technology Data Exchange (ETDEWEB)

    Morris, C.J.; Doctor, S.R.

    1977-12-01

    This report documents an independent review and evaluation of sodium leak detection methods described in the Clinch River Breeder Reactor Preliminary Safety Analysis Report. Only information in publicly available documents was used in making the assessments.

  6. Health physics experiences in the operation of Fast Breeder Test Reactor (FBTR)

    International Nuclear Information System (INIS)

    This paper presents the health physics experience gained with the operation of the Fast Breeder Test Reactor (FBTR), which was made critical in October 1985. Major operations that were carried out and the associated health physics surveillance are highlighted. (author)

  7. Development of Genetic Algorithm based Neural Network model for parameter estimation of Fast Breeder Reactor Subsystem

    OpenAIRE

    Subhra Rani Patra; R. Jehadeesan; Rajeswari, S.

    2012-01-01

    This work provides the construction of Genetic Algorithm based Neural Network for parameter estimation of Fast Breeder Test Reactor (FBTR) Subsystem. The parameter estimated here is temperature of Intermediate Heat Exchanger of Fast Breeder Test Reactor. Genetic Algorithm based Neural Network is a global search algorithm having less probability of being trapped in local minimum problem as compared to Standard Back Propagation algorithm which is a local search algorithm. The various developmen...

  8. Decay Heat Removal for the Liquid Metal Fast Breeder Reactor

    International Nuclear Information System (INIS)

    The functional and reliability requirements of the decay heat removal systems are described. The reliability requirement and its rationale as adequate assurance that public health and safety are safeguarded are presented. The means by which the reliability of the decay heat removal systems are established to meet their requirement are identified. The heat removal systems and their operating characteristics are described. The discussion includes the overflow heat removal service and its role in decay heat removal if needed. The details of the systems are described to demonstrate the elements of redundancy and diversity in the systems design. The quantitative reliability assessment is presented, including the reliability model, the most important assumptions on which the analysis is based, sources of failure data, and the preliminary numerical results. Finally, the qualitative analyses and administrative controls will be discussed which ensure reliability attainment in design, fabrication, and operation, including minimization of common mode failures. A component test program is planned to provide reliability data on selected critical heat removal system equipment. This test plan is described including a definition of the test parameters of greatest interest and the motivation for the test article selection. A long range plan is also in place to collect plant operational data and the broad outlines of this plan are described. The paper closes with a statement of the high reliability of the Clinch River Breeder Reactor Plant decay heat removal systems and a summary of the supporting arguments. (author)

  9. Development of chemical sensors for Fast Breeder Reactor Technology

    International Nuclear Information System (INIS)

    Fast breeder reactors use liquid sodium as heat transfer medium and generate high pressure steam at the steam generator to run the turbine. This high pressure steam is separated from sodium coolant by ferritic steel tubes of 4 to 5 mm wall thickness. Development of any material defect in these heat exchanger tubes during their service would result in the ingress of high pressure steam into the sodium circuit leading to sodium-water reactions. A high temperature electrochemical hydrogen sensor based on CaBr2-CaHBr solid electrolyte and capable of measuring ppb levels of dissolved hydrogen in sodium has been developed at the laboratory. A very sensitive system, using thermal conductivity detector and semiconducting oxide based sensor has also been developed for continuous monitoring of hydrogen levels in argon cover gas. An electrochemical carbon sensor using a molten carbonate electrolyte and an oxygen sensor based on yttria doped thoria oxide electrolyte are also under advanced stage of development for measuring carbon and oxygen levels in sodium. Materials chemistry issues involved in developing these sensors and their operational experience in sodium system are highlighted in this presentation

  10. Feasibility studies on commercialized fast breeder reactor cycle system

    International Nuclear Information System (INIS)

    JNC (Japan Nuclear Cycle Development Institute) and the electric utilities in Japan have established a new organization to develop a commercialized fast breeder reactor (FBR) cycle system since July 1, 1999, feasibility studies (F/S) have been undertaken in order to determine the promising concepts and to define the necessary R and D tasks. In the first two-year phase, a number of candidate concepts will be selected from various options, featuring innovative technologies. In the F/S, the options are evaluated and conceptual designs are examined considering the attainable perspectives for following: 1) ensuring safety, 2) economic competitiveness to future LWRs, 3) efficient utilization of resources, 4) reduction of environmental burden and 5) enhancement of nuclear non-proliferation. The F/S should also guide the necessary R and D to commercialize FBR cycle system. In particular enhanced technologies should be integrated in order to ensure nuclear non-proliferation. In the second five-year phase of the F/S, scaled engineering tests will be conducted. Based on the test data, a comprehensive evaluation will be conducted to confirm the technical attainability of candidate concepts. A few proposals for the commercialization of the FBR cycle system will be proposed. (author)

  11. Method of advancing research and development of fast breeder reactors

    International Nuclear Information System (INIS)

    In the long term plan of atomic energy development and utilization, fast breeder reactors are to be developed as the main of the future nuclear power generation in Japan, and when their development is advanced, it has been decided to positively aim at building up the plutonium utilization system using FBRs superior to the uranium utilization system using LWRs. Also it has been decided that the development of FBRs requires to exert incessant efforts for a considerable long period under the proper cooperation system of government and people, and as for its concrete development, hereafter the deliberation is to be carried out in succession by the expert subcommittee on FBR development projects of the Atomic Energy Commission. The subcommittee was founded in May, 1986, to deliberate on the long term promotion measures for FBR development, the measures for promoting the research and development, the examination of the basic specification of a demonstration FBR, the measures for promoting international cooperation, and other important matters. As the results of investigation, the situation around the development of FBRs, the fundamentals at the time of promoting the research and development, the subjects of the research and development and so on are reported. (Kako, I.)

  12. Thermodynamics of ceramic breeder materials for fusion reactors

    International Nuclear Information System (INIS)

    Based on known or deduced phase relationships in ternary lithium oxygen systems such as Li-Al-O, Li-Si-O and Li-Zr-O, the unknown free enthalpy of formation values of ternary compounds are calculated starting from the known data of the compounds of the binary border systems. Criterion for the data assessment is interconsistency of the data of all the compounds within a given multi-component system. With the help of these data the development of partial pressures during the breeding process can be calculated for all the compounds of interest. In order to facilitate a compatibility assessment the quaternary systems Cr-Li-Si-O, Fe-Li-Si-O and Be-Li-Si-O were also investigated and thermodynamic data of pertinent ternary and quaternary compounds determined. Both the partial pressure development and the compatibility behaviour of a lithium containing compound are criteria for its qualification as a breeder material for a fusion reactor. (orig.)

  13. Defect assessment procedure: A french approach for fast breeder reactors

    International Nuclear Information System (INIS)

    As a result of a collaborative effort between Commissariat a l'Energie Atomique, Electricite de France, and NOVATOME to produce and improve rules for fast breeder reactors, RCC-MR, an interim defect assessment procedure is now available in the first draft version (appendix A16). This procedure addresses defects detected during in-service inspection for reactor components operating at moderate or high temperature conditions. Three stages have been considered: initiation, propagation under cyclic loading with or without holdtime and crack instability by ductile and creep rupture. For each of these topics, procedures and rules based on fracture mechanics are proposed. Prediction of initiation is obtained by a simplified method named σd method which relies on the evaluation of the real stress-strain history on a small distance d (d = 0.05 mm for 316L(N) austenitic steel) close to the crack front and material characteristics (limiting stresses) that are available in nuclear codes. This method has been developed for fatigue, creep and creep-fatigue conditions. Defect growth assessment is performed for fatigue and creep-fatigue conditions. For creep-fatigue conditions, fatigue and creep crack growth per cycle are calculated separately and the total crack extension is taken as the sum of the two contributions. Extensive use of simplified method for estimating J (Js method) is made and developed when mechanical and thermal loadings are specified. On the final defect size, assessment may be made in order to avoid crack instability by ductile and creep rupture and collapse load on the remaining. The organization and contents of the present version of this appendix A16 is described. An overview of each specific rule is given

  14. Status and prospects of advanced fissile fuel breeders

    International Nuclear Information System (INIS)

    Fusion--fission hybrid systems, fast breeder systems, and accelerator breeder systems were compared on a common basis using a simple economic model. Electricity prices based on system capital costs only were computed, and were plotted as functions of five key breeder system parameters. Nominally, hybrid system electricity costs were about twenty-five percent lower than fast breeder system electricity costs, and fast breeder system electricity costs were about forty percent lower than accelerator breeder system electricity costs. In addition, hybrid system electricity costs were very insensitive to key parameter variations on the average, fast breeder system electricity costs were moderately sensitive to key parameter variations on the average, and accelerator breeder system electricity costs were the most sensitive to key parameter variations on the average

  15. Fast Breeder Reactor Development in France During 1987

    International Nuclear Information System (INIS)

    On March 8, 1987, a ''sodium leak'' alarm signal was received in the Creys-Malville control room. By the end of March, it had been established beyond all doubt that sodium was, in fact, leaking into the fuel storage drum inter-vessel gap. The reactor has been shut down since May 26. The origin of the leak was located on September 5, after complete drainage of the main tank. Despite the fact that the leak was confined, had had no radiological consequences and cast no doubts on the safe operation of the reactor, the impact of this incident on public opinion, both in France and in the neighbouring European states, was considerable. Two facts would appear to have been decisive. The first was that the reactor had not been shut down immediately, the second was that the leak was only detected and localized in September: it was difficult for people to understand that before its exact position could be determined, certain operations (transfer of a few subassemblies to the reactor core, unloading of the fuel storage drum) had to be performed

  16. Analysis of the running-in phase of a Passively Safe Thorium Breeder Pebble Bed Reactor

    International Nuclear Information System (INIS)

    Highlights: • This work analyzes important trends of the running-in phase of a thorium breeder PBR. • Depletion equations are solved for important actinides and a fission product pair. • Breeding U-233 is achieved in 7 years by cleverly adjusting the feed fuel enrichment. • A safety analysis shows the thorium PBR is passively safe during the running-in phase. - Abstract: The present work investigates the running-in phase of a 100 MWth Passively Safe Thorium Breeder Pebble Bed Reactor (PBR), a conceptual design introduced in previous equilibrium core design studies by the authors. Since U-233 is not available in nature, an alternative fuel, e.g. U-235/U-238, is required to start such a reactor. This work investigates how long it takes to converge to the equilibrium core composition and to achieve a net production of U-233, and how this can be accelerated. For this purpose, a fast and flexible calculation scheme was developed to analyze these aspects of the running-in phase. Depletion equations with an axial fuel movement term are solved in MATLAB for the most relevant actinides (Th-232, Pa-233, U-233, U-234, U-235, U-236 and U-238) and the fission products are lumped into a fission product pair. A finite difference discretization is used for the axial coordinate in combination with an implicit Euler time discretization scheme. Results show that a time dependent adjustment scheme for the enrichment (in case of U-235/U-238 start-up fuel) or U-233 weight fraction of the feed driver fuel helps to restrict excess reactivity, to improve the fuel economy and to achieve a net production of U-233 faster. After using U-235/U-238 startup fuel for 1300 days, the system starts to work as a breeder, i.e. the U-233 (and Pa-233) extraction rate exceeds the U-233 feed rate, within 7 years after start of reactor operation. The final part of the work presents a basic safety analysis, which shows that the thorium PBR fulfills the same passive safety requirements as the

  17. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    A nuclear reactor fuel element comprising a column of vibration compacted fuel which is retained in consolidated condition by a thimble shaped plug. The plug is wedged into gripping engagement with the wall of the sheath by a wedge. The wedge material has a lower coefficient of expansion than the sheath material so that at reactor operating temperature the retainer can relax sufficient to accommodate thermal expansion of the column of fuel. (author)

  18. Evaluation of symbiotic energy system between gas-cooled fast breeder reactor (GCFR) and multi-purpose very high temperature reactor (VHTR), (4)

    International Nuclear Information System (INIS)

    The conceptual design study of 1000 MWe gas-cooled fast breeder reactor (GCFR), which is used in the GCFR-VHTR symbiotic energy system, has been performed. In this report, the transient response of the GCFR core to accident events has been analyzed and safety performance of the 1000 MWe GCFR has been evaluated considering the analyses. A depressurization accident caused by failure of a primary coolant system and a reactivity insertion accident due to withdrawal of a control rod have been analyzed using nuclear and thermo-hydraulic coupled program MR-X developed for kinetics analysis of gas-cooled fast breeder reactors. The maximum fuel and cladding temperatures are most important problem to be analysed during a trangient of a gas-cooled fast breeder reactors. The analyses show that reliable reactor shutdown and emergency cooling systems are most important to achieve successful cold shutdown well before leading to core damage and also that no severe failures of fuel pin and cladding occures by working above mentioned safety systems well during the accidents. (author)

  19. The present status of the fast breeder reactor industrialization in western Europe

    International Nuclear Information System (INIS)

    The development of the liquid metal fast breeder reactor in Europe started in the mid-fifties, after the successful operation of EBR-1 at ARCO, Idaho, in 1951. A more and more integrated development among the countries of the European Community culminated in 1986 with the beginning to power of the 1200 MWe SUPERPHENIX plant at Creys-Malville, France. The road is now open towards the full industrialization of the liquid metal fast breeder reactor at the moment, in 2005, when the first European thermal neutron power reactor station will have to be decommissioned and replaced. The European programme aims at providing the utilities at that time with a clear choice between thermal neutron reactors and fast breeder reactors, both economical but very different in their use of the limited natural resource that uranium is. (author)

  20. Present status of the fast breeder reactor industrialization in western Europe

    International Nuclear Information System (INIS)

    The development of the liquid metal fast breeder reactor in Europe started in the mid-fifties, after the successful operation of EBR-1 at ARCO, Idaho, in 1951. A more and more integrated development among the countries of the European Community culminated in 1986 with the startup of the 1200 MWe SUPERPHENIX plant at Creys-Malville, France. The road is now open towards the full industrialization of the liquid metal fast breeder reactor at the moment, in 2005, when the first European thermal neutron power reactor station will have to be decommissioned and replaced. The European programme aims at providing the utilities at that time with a clear choice between thermal neutron reactors and fast breeder reactors, both economical but very different in their use of the limited natural resources that uranium

  1. C-scope under-sodium viewer for sodium-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    A C-scope under-sodium viewer has been developed for monitoring the interior of sodium-cooled fast breeder reactors. Consisting of a transducer that emits and receives ultrasonic waves under liquid sodium, a mechanism that drives the transducer under liquid sodium and an image displaying section, it inspects the fuel assembly through its image in optically opaque high-temperature (3000C) liquid sodium. The results of its evaluation test are: (1) The transducer could continue satisfactory operation under 3500C (at the highest) sodium for more than a month. (2) The driving mechanism, though it was the first of the kind appearing in Japan, has been proved that it could continue operation for a week under 3000C sodium. (3) The image displaying section, in spite of the low speed of the transducer (below 20 rpm), could display stable and clear images. (4) The image in 3000C was as clear as that in room-temperature water. (auth.)

  2. Analysis of passive shutdown capability for a loss of flow accident in a medium sized liquid metal fast breeder reactor

    International Nuclear Information System (INIS)

    The passive shutdown capability of a medium sized (500 MWe) liquid-metal fast breeder reactor with oxide, carbide and metal fuels has been analysed for a loss of flow accident using static and dynamic analysis methods. The carbide fuel is assumed to be He-bonded as well as Na-bonded. The dependence of the passive safety on the flow halving time constant of the loss of flow incident and the feedback components, like radial core expansion due to subassembly spacer pad heating and differential control rod expansion due to heating of the control rod suspension mechanism, is highlighted. (author)

  3. Gas-cooled fast breeder reactor shielding benchmark calculation

    Energy Technology Data Exchange (ETDEWEB)

    Rouse, C.A.; Mathews, D.R.; Koch, P.K.

    1977-01-01

    This report summarizes the results of a shielding benchmark calculation performed by General Atomic (GA) and Oak Ridge National Laboratory (ORNL). The problem analyzed was a neutron-coupled gamma ray transport calculation of the core blanket shield of the 300-MW(e) gas-cooled fast breeder reactor (GCFR). Comparison of the initial GA and ORNL results indicated good agreement for fast fluxes (E greater than 0.9 MeV and E greater than 0.086 MeV) but poor agreement for epithermal and thermal neutron fluxes. Examination of the results revealed that a deficiency in the GA fine-group cross section preparation code was responsible for the differences in the GA and ORNL iron cross sections. Modification of the GA cross sections to include self-shielding was accomplished, and the updated GA benchmark calculation performed with the self-shielded iron cross sections was in excellent agreement with the ORNL results for fast neutron fluxes with E greater than 0.9 MeV and E greater than 0.086 MeV and in good agreement for epithermal and thermal fluxes. The agreement of the gamma heating rates also improved significantly. Thus, it was concluded that the good agreement of the GA and ORNL neutron-coupled gamma ray transport calculation indicates that (1) the methods and cross sections used by both laboratories were compatible and consistent and (2) the use of 24 neutron energy groups and 15 gamma energy groups by GA was adequate compared with the use of 51 neutron energy groups and 25 gamma energy groups by ORNL.

  4. Network Representation of Design Knowledge of Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    A method of design knowledge representation was studied for the Japanese fast breeder reactor Monju, aiming at enhanced understanding of engineering considerations with mutual relations. Taking over design knowledge of Monju to next generation designers/engineers to be in charge of design of future FRs is by no means easy, in contrast with operation and maintenance knowledge which can be acquired in the real plant operation and maintenance. Specifications of the as-is Monju contains only a small part of the entire design knowledge, mainly by two reasons. Firstly, reasons for selecting the as-is specifications can not be understood until reaching proper knowledge source. Secondly, there are many rejected options on the design specifications. Design specifications are selected along with technical dependencies among a huge number and diversified specification items. Decisions design are made basically along with these dependencies which can hardly be traced in the currently available database or document libraries. Reasons for the rejections of options need to be profoundly understood, because those are not certainly due to technical inferiority. Some of rejected options can be worth reconsidering in the future, possibly by technical advances in materials, high-precision prediction software tools, rationalized standards/code, etc. The authors propose a new design knowledge representation approach based on networking of knowledge nodes along with the mutual dependencies. A prototype software has been developed and a basic performance test was made to visualize the dependency network. An additional function to enable design case studies on hypothetical adoptions of rejected options is now under consideration. (author)

  5. Argon entrainment into liquid sodium in fast breeder reactor

    International Nuclear Information System (INIS)

    Highlights: ► In the present work, different types of entrainment mechanisms have been studied. ► Onset of entrainment could be characterized with turbulent intensities. ► For vortex type entrainment, a correlation has been proposed. - Abstract: Gas entrainment in pool type sodium cooled fast breeder reactors has been a subject of great interest for a quite long time now. The issue of entrainment of argon cover gas in LMFBR's is being addressed by fundamental studies. Present work focuses on characterization of onset of shear type entrainment and liquid fall type entrainment based on mean velocity and turbulent kinetic energy at liquid surface. Study also includes characterization of onset of vortex type entrainment based on mean velocities (time averaged) in the outlet pipe. Experiments were carried out to characterize shear type entrainment in stirred tank with different impeller geometries with air–water and xylene–water systems. Onset of liquid fall type entrainment was studied with cylindrical tank with a nozzle whose input angle varied. Mean and r.m.s. velocity profiles near the liquid surface were measured with the help of ultrasonic velocity profiler (UVP). The results are compared with other literature. It is observed that the onset of entrainment can be characterized by the turbulent kinetic energy near the free liquid surface. Re-submergence angle was measured and r.m.s. velocities found to be in the same range as in case of shear type of entrainment. Cylindrical tank with tangential inlet and bottom outlet was used to study onset of vortex formation. Effect of different parameters like outlet diameter, tank diameter and liquid height in the tank on critical velocity was studied and correlation has been proposed.

  6. Progress report on fast breeder reactor development in Japan

    International Nuclear Information System (INIS)

    As for the experimental fast reactor ''Joyo'', the power increase test has been carried out since April, and the power output was raised stepwise up to 40 MW. The power output, core behavior, plant characteristics as well as shielding integrity were measured at each power level. The examination for licensing the power increase to 75 and 100 MW is still continued by the Committee No. 130. The preparation of various codes required for the core characteristic analysis is in progress. As for the development of the prototype fast reactor ''Monju'', the Construction Preliminary Design (1) was evaluated, and the studies on the specifications of the Construction Preliminary Design (2) are carried out. In respect to the analysis for the Safety Licensing, the analysis of decay heat, the development of an analytical code regarding the rupture propagation in heat transfer tubes for steam generators and others are under way. Technological investigation is carried out to obtain the overseas informations on the safety standards for FBRs and LMFBR technologies. The technical specifications for the preliminary design of the demonstration fast reactor are being prepared. The researches and developments of reactor physics, the structural components of ''Joyo'' and ''monju'', instrumentation and control, sodium technology, fuel materials, structural materials, safety and steam generators are reported, respectively. (Kako, I.)

  7. Zeolite membranes and palladium membrane reactor for tritium extraction from the breeder blankets of ITER and DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Demange, D., E-mail: david.demange@kit.edu; Borisevich, O.; Gramlich, N.; Wagner, R.; Welte, S.

    2013-10-15

    Highlights: • We present a new concept to recover tritium from the helium in breeder blankets. • Zeolite membranes are fully tritium compatible and can pre-concentrate tritiated molecules. • PERMCAT catalytic membrane reactor recovers tritium to be reused in the fuel cycle. -- Abstract: While the tritium technology for the inner DT fuel cycle of fusion reactors shall be demonstrated in ITER, the tritium management in the breeder blanket remains very challenging. Most of the process options rely on ad(b)sorption/desorption cycles, using dedicated packed beds to handle separately the molecular and oxide forms of tritium. This approach seems satisfactory for ITER, but seems difficult to scale up to DEMO. The alternative use of a catalytic membrane reactor in combination with inorganic membranes would simplify and improve the overall tritium management. Zeolite membranes should enable in a single step the pre-concentration of all tritiated species. This tritium enriched stream could be afterwards processed using PERMCAT (catalytic Pd-based membrane reactor) to finally recover the tritium in its pure molecular form. This paper discusses at the conceptual level such approach. The latest experimental results on zeolite membrane and multi-tube PERMCAT reactor are presented. Next R and D activities for technical scale demonstrations and refined simulation tools are proposed to finally estimate the sizes of the components to be operated in ITER and DEMO.

  8. Development of high nitrogen electrodes for fast breeder reactor applications

    International Nuclear Information System (INIS)

    Austenitic stainless steels of AISI type 316 (316 SS) and its variants are used extensively as structural material for the components of fast reactors operating at temperature up to 823 K. SS 316LN has been chosen as the major structural material for the construction of Prototype Fast Breeder Reactor (PFBR) with a targeted service life of 40 years. To reduce the risk of sensitization in SS 316LN, the carbon content has been reduced to less than 0.03 wt%, and the nitrogen content has been specified as 0.08 wt% to compensate the loss in strength due to the reduced carbon content. An improved version of this alloy with nitrogen content of 0.14 wt% in a frilly austenite matrix has been developed for the future FBRs, to enhance the service life of the structural components up to 60 years. Indigenously developed modified E3 16-1 5 electrodes were used for the fabrication of PFBR components to enhance the structural reliability of the components. The modifications from AWS/ASME SFA 5.4 include stringent composition limits, narrow range of ferrite content, and impact toughness after aging at 1023K for 100h, tensile properties at elevated (service) temperatures and intergranular corrosion (IGC) test as per ASTM A262 Practice E. Since the improved version alloy is rich in nitrogen content than the existing alloy, it has become necessary to develop a welding consumable with reasonably good weldability that is suitable for the fabrication of future FBR components. At present there are no commercially available welding consumables to weld these steels and the development is under way. In this work, a matching consumable methodology was adopted to develop the welding consumable. However, as per specification targeting the chemistry, solidification mode and delta ferrite was challenging, since the solidification mode of the weld metal shifts to fully austenitic region due to dilution of nitrogen from the base metal, which may increase the risk of hot cracking susceptibility

  9. Fast breeder reactors insertion in a D2O - natural U nuclear power plants park

    International Nuclear Information System (INIS)

    A model for the evolution of Argentine's installed nuclear power for the next 40 years is presented. The consequences of fast breeder reactors' introduction are studied in both autarchic Pu cycle and a limited reprocessing system. The passage of a reactor park like the national, of natural U - heavy water to one of fast breeder reactors, can only be obtained in a very long term due, fundamentally, to the need of Pu produced for those to feed the last ones. (M.E.L.)

  10. Structural analysis of reactor fuel elements

    International Nuclear Information System (INIS)

    An overview of fuel-element modeling is presented that traces the development of codes for the prediction of light-water-reactor and fast-breeder-reactor fuel-element performance. It is concluded that although the mathematical analysis is now far advanced, the development and incorporation of mechanistic constitutive equations has not kept pace. The resultant reliance on empirical correlations severely limits the physical insight that can be gained from code extrapolations. Current efforts include modeling of alternate fuel systems, analysis of local fuel-cladding interactions, and development of a predictive capability for off-normal behavior. Future work should help remedy the current constitutive deficiencies and should include the development of deterministic failure criteria for use in design

  11. Y2K issues for real time computer systems for fast breeder test reactor

    International Nuclear Information System (INIS)

    Presentation shows the classification of real time systems related to operation, control and monitoring of the fast breeder test reactor. Software life cycle includes software requirement specification, software design description, coding, commissioning, operation and management. A software scheme in supervisory computer of fast breeder test rector is described with the twenty years of experience in design, development, installation, commissioning, operation and maintenance of computer based supervision control system for nuclear installation with a particular emphasis on solving the Y2K problem

  12. Approaches to measurement of thermal-hydraulic parameters in liquid-metal-cooled fast breeder reactors

    International Nuclear Information System (INIS)

    This lecture considers instrumentation for liquid-metal-cooled fast breeder reactors (LMFBR's). Included is instrumentation to measure sodium flow, pressure, temperature, acoustic noise, and sodium purity. It is divided into three major parts: (1) measurement requirements for sodium cooled reactor systems, (2) in-core and out-of-core measurements in liquid metal systems, and (3) performance measurements of water steam generators

  13. Conceptual design of a pool type molten salt breeder reactor

    International Nuclear Information System (INIS)

    The renewed interest in molten salt coolant technology is backed by the 50 years history of molten salt nuclear technology development, mainly in Oak Ridge National Laboratory (ORNL). In Indian context MSBR is found to be one of the options for sustainable nuclear energy generation, especially in the third stage of the nuclear programme. The system can be operated at high temperature which makes high efficiency power conversion and efficient hydrogen generation through thermo-chemical reactions possible. At present development is in progress in BARC on two molten salt reactor concepts, one is pool type and the other is loop type. Here the design of pool type concept with 850MWe power is described. The core is designed to operate in the fast spectrum region so the conversion of 233U breeding is possible from thorium. Preliminary thermal hydraulic analysis is carried out with LiF-ThF4-UF4 as the primary fuel and coolant. The blanket material is also a molten salt, LiF-ThF4. Reactor physics calculations are also carried out for the feasibility studies of the core design of the reactor. FLiNaK is used as the secondary coolant for the calculations. Both forced circulation and natural circulation options are evaluated. (author)

  14. 03 - Sodium cooled fast breeder fourth-generation reactors - The technological demonstrator ASTRID

    International Nuclear Information System (INIS)

    After a discussion of the past experience gained on fast breeder reactors in the world (benefits, difficulties and problematics), the authors discuss the main improvement domains and the associated R and D advances (reactor safety, prevention and mitigation of severe accidents, the sodium-water risk, detection of sodium leaks, increased availability, instrumentation and inspection, control and repairability, assembly handling and washing). Then, they describe the technical requirements and safety objectives of the ASTRID experimental project, notably with its reactivity management, cooling management, and radiological containment management functions. They describe and discuss requirements to be met and choices made for Astrid, and the design options for its various components (core and fuels, nuclear heater, energy conversion system, fuel assembly handling, instrumentation and in-service inspection, control and command). They present the installations which are associated with the ASTRID cycle, evoke the development and use of simulations and codes, describe the industrial organization and the international collaboration about the ASTRID project, present the planning and cost definition

  15. Modeling and analysis of the unprotected loss-of-flow accident in the Clinch River Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Morris, E.E.; Dunn, F.E.; Simms, R.; Gruber, E.E.

    1985-01-01

    The influence of fission-gas-driven fuel compaction on the energetics resulting from a loss-of-flow accident was estimated with the aid of the SAS3D accident analysis code. The analysis was carried out as part of the Clinch River Breeder Reactor licensing process. The TREAT tests L6, L7, and R8 were analyzed to assist in the modeling of fuel motion and the effects of plenum fission-gas release on coolant and clad dynamics. Special, conservative modeling was introduced to evaluate the effect of fission-gas pressure on the motion of the upper fuel pin segment following disruption. For the nominal sodium-void worth, fission-gas-driven fuel compaction did not adversely affect the outcome of the transient. When uncertainties in the sodium-void worth were considered, however, it was found that if fuel compaction occurs, loss-of-flow driven transient overpower phenomenology could not be precluded.

  16. Design and fabrication of steam generators (superheaters) for the prototype fast breeder reactor 'MONJU'

    International Nuclear Information System (INIS)

    In liquid metal-cooled fast breeder reactors, steam generators are one of the important equipments, and emphasis has been placed on their development in various countries in the world. Also in Japan, centering around the Power Reactor and Nuclear Fuel Development Corp., the research and development in the wide range from the fundamentals on heat transfer and flow, materials and strength for steam generators to the manufacture, operation and various tests of large mock-ups including a 50 MW steam generator have been carried out. Further, as for the manufacture and inspection, the improvement of the method of welding tubes and tube plates, the adoption of a fine focus X-ray inspection apparatus and others were carried out. Moreover, as the maintenance technique, the ultrasonic flaw detection probes for the heating tubes were developed. The steam generators (superheaters) for the FBR 'Monju' power station are the heat exchangers of helical coil tube-shell type using SUS 321 steel as the heating tube material. Based on the results of these research and development, the design and manufacture of these superheaters and their installation in the reactor auxiliary building of the FBR 'Monju' power station were completed. The outline of the design, the research and development and the manufacture of the steam generators (superheaters) are reported. (K.I.)

  17. Phase 1 of feasibility studies on commercialized fast breeder reactor cycle system

    Energy Technology Data Exchange (ETDEWEB)

    Konomura, Mamoru [Japan Nuclear Cycle Development Inst., Tokai, Ibaraki (Japan)

    2001-09-01

    The situation of the feasibility study on commercialized fast breeder reactor cycle system is reported. This is the joint study of JNC and major domestic electric power companies. The study intends to construct and propose the concept of candidate FBR cycle responding various needs in 21 century. The target of the study is to extract plural candidates of commercialized FBR cycle system including FBR plant, reprocessing facility and fuel fabrication facility and to propose realization scenario such as R and D schedule. The study settles on the comprehensive evaluation guideline on not only safety but also on economy, effective utilization of resources, reduction of environmental load and non-proliferation of nuclear materials. Based on the guideline, the design target items are established. During the Phase 1 (FY1999 and 2000), the extensive FBR systems are surveyed in the study. As coolant, Na heavy metals such as Pb, PbBi, gases such as CO2, He and water (H2O and D2O) are surveyed. As fuel, pin type such as oxide, metal and nitride, coated particle type of oxide and nitride and liquid fuel are considered combined with coolant. The final report of Phase 1 will be published shortly. In Phase 2 of the study starting FY2001 until FY2005, the extraction of the candidates will be performed. (K. Tsuchihashi)

  18. Phase 1 of feasibility studies on commercialized fast breeder reactor cycle system

    International Nuclear Information System (INIS)

    The situation of the feasibility study on commercialized fast breeder reactor cycle system is reported. This is the joint study of JNC and major domestic electric power companies. The study intends to construct and propose the concept of candidate FBR cycle responding various needs in 21 century. The target of the study is to extract plural candidates of commercialized FBR cycle system including FBR plant, reprocessing facility and fuel fabrication facility and to propose realization scenario such as R and D schedule. The study settles on the comprehensive evaluation guideline on not only safety but also on economy, effective utilization of resources, reduction of environmental load and non-proliferation of nuclear materials. Based on the guideline, the design target items are established. During the Phase 1 (FY1999 and 2000), the extensive FBR systems are surveyed in the study. As coolant, Na heavy metals such as Pb, PbBi, gases such as CO2, He and water (H2O and D2O) are surveyed. As fuel, pin type such as oxide, metal and nitride, coated particle type of oxide and nitride and liquid fuel are considered combined with coolant. The final report of Phase 1 will be published shortly. In Phase 2 of the study starting FY2001 until FY2005, the extraction of the candidates will be performed. (K. Tsuchihashi)

  19. History of fast reactor fuel development

    Science.gov (United States)

    Kittel, J. H.; Frost, B. R. T.; Mustelier, J. P.; Bagley, K. Q.; Crittenden, G. C.; Van Dievoet, J.

    1993-09-01

    The first fast breeder reactors, constructed in the 1945-1960 time period, used metallic fuels composed of uranium, plutonium, or their alloys. They were chosen because most existing reactor operating experience had been obtained on metallic fuels and because they provided the highest breeding ratios. Difficulties in obtaining adequate dimensional stability in metallic fuel elements under conditions of high fuel burnup led in the 1960s to the virtual worldwide choice of ceramic fuels. Although ceramic fuels provide lower breeding performance, this objective is no longer an important consideration in most national programs. Mixed uranium and plutonium dioxide became the ceramic fuel that has received the widest use. The more advanced ceramic fuels, mixed uranium and plutonium carbides and nitrides, continue under development. More recently, metal fuel elements of improved design have joined ceramic fuels in achieving goal burnups of 15 to 20 percent. Low-swelling fuel cladding alloys have also been continuously developed to deal with the unexpected problem of void formation in stainless steels subjected to fast neutron irradiation, a phenomenon first observed in the 1960s.

  20. Cladding and wrapper development for fast breeder reactor high performance

    International Nuclear Information System (INIS)

    In order to ensure economic performance, of both the existing reactors and the future EFR, much recent research has been carried out within the framework of the European R and D agreement to examine the properties of various wrapper and cladding alloys. This paper reviews the status of the European research and development programmes on these steels and highlights the most striking results. For the cladding alloys, results on dimensional stability and tensile properties for fuel pin cladding irradiated in PFR or Phenix will be given. As for wrappers the presently available results of those wrappers irradiated in Phenix and PFR show that both ferritic steels are very good candidates and that on the basis of our present knowledge most of the properties are satisfactory for wrapper applications

  1. Cermet fuel reactors

    International Nuclear Information System (INIS)

    Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. The concept evolved in the 1960's with the objective of developing a reactor design which could be used for a wide range of mobile power generation systems including both Brayton and Rankine power conversion cycles. High temperature thermal cycling tests and in-reactor irradiation tests using cermet fuel were carried out by General Electric in the 1960's as part of the 710 Development Program and by Argonne National laboratory in a subsequent activity. Cermet fuel development programs are currently underway at Argonne National laboratory and Pacific Northwest Laboratory as part of the Multi-Megawatt Space Power Program. Key features of the cermet fueled reactor design are 1) the ability to achieve very high coolant exit temperatures, and 2) thermal shock resistance during rapid power changes, and 3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, there is a potential for achieving a long operating life because of 1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and 2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core

  2. The Future of Nuclear Energy: Facts and Fiction Chapter IV: Energy from Breeder Reactors and from Fusion?

    OpenAIRE

    Dittmar, Michael

    2009-01-01

    The accumulated knowledge and the prospects for commercial energy production from fission breeder and fusion reactors are analyzed in this report. The publicly available data from past experimental breeder reactors indicate that a large number of unsolved technological problems exist and that the amount of "created" fissile material, either from the U238 --> Pu239 or from the Th232 --> U233 cycle, is still far below the breeder requirements and optimistic theoretical expectations. Thus huge e...

  3. Operation and maintenance experience with control rod and their drive mechanisms of fast breeder test reactor

    International Nuclear Information System (INIS)

    This paper explains the functional and construction features of Control Rod Drive Mechanism (CRDM) and control rod used in Fast Breeder Test Reactor (FBTR) which is a 40 MWt loop type sodium cooled fast reactor. It discusses all safety related incidents and failures encountered during its service in reactor, the solutions evolved and modifications carried out to prevent recurrence. It also details the maintenance activities and periodical surveillance carried out. The results of a reliability analysis done are also discussed. (author)

  4. Super-Phenix: first step in the commercial development of breeder reactors

    International Nuclear Information System (INIS)

    Super-Phenix is an indispensable step in a program for the development of the breeder reactors. Its power and technical options are recalled. The administrative directives -particularly concerning safety- are in agreement with the regulation enforced in France relative to nuclear power stations and work has begun on the Creys-Malville site within the international framework where the development of breeders in Europe now finds its place. The saving in uranium due to breeders is certain, but their cost prices remain to be proved. The construction of Super-Phenix is the proof that it is possible to rely on the breeders in the final energetic results for the future, but it is of a type which can still be modified to a considerable extent, especially in order to reduce investments

  5. Status and prospects of thermal breeders

    International Nuclear Information System (INIS)

    The main objective of this cooperative study and of this report is to evaluate the extent to which thermal breeders might complement or serve as an alternative to fast breeders in solving the long-term nuclear fuel supply problem. A secondary objective is to consider in a general way issues such as proliferation, safety, environmental impacts, economics, power plant availability, and fuel cycle versatility to determine whether thermal breeder reactors offer advantages or disadvantages with respect to such issues

  6. Quality assurance in technology development for The Clinch River Breeder Reactor Plant Project

    International Nuclear Information System (INIS)

    The Clinch River Breeder Reactor Plant Project is the nation's first large-scale demonstration of the Liquid Metal Fast Breeder Reactor (LMFBR) concept. The Project has established an overall program of plans and actions to assure that the plant will perform as required. The program has been established and is being implemented in accordance with Department of Energy Standard RDT F 2-2. It is being applied to all parts of the plant, including the development of technology supporting its design and licensing activity. A discussion of the program as it is applied to development is presented

  7. Storage and Management of Spent Fuel of Fast Reactors in India

    International Nuclear Information System (INIS)

    Deploying fast breeder reactors on a commercial scale is vital for India in order to utilise the vast thorium reserves and to meet the long term energy needs of the country. As a first step, a 40 MWt, 13.5 MWe Fast Breeder Test Reactor (FBTR) was constructed and has been in operation since 1985. FBTR utilises mixed carbide fuel. The spent fuel sub assembly of FBTR is stored in an argon inverted container and cooled by air. India took the construction of indigenously designed 500 MWe Prototype Fast Breeder Reactor (PFBR) in September 2003. The reactor is in advanced stage of construction. It is planned to construct six more FBRs of 500 MWe each by the year 2023. All the 500 MWe fast reactors utilise oxide fuel, the most proven fuel in the case of the fast reactors. The spent fuel subassembly after internal storage for one campaign of 8 calendar months is washed to remove sodium and stored in water in a spent fuel storage lined with SS 304L liner and with a provision of leak monitoring of the liner welds. The failed subassembly is stored in a leak tight container before putting in the spent fuel storage. The road to getting faster growth in Indian indigenous nuclear power programme is through use of metallic fuel in fast breeder reactors. The metallic fuel design gives high breeding ratio and thus shorter doubling time. Both sodium and mechanical bonded metallic fuel design are under development. The growth of fast reactors is sensitive to the time period the spent fuel is out of the reactor i.e. from the time spent fuel is out of the reactor, cooled, reprocessed, refabricated and put back into the reactor. Storage of spent fuel subassemblies in sodium and co-located pyro-reprocessing facility is planned for the metallic fuel reactors. The paper describes in brief the system for storage and management of spent fuel of FBTR, PFBR and metallic fuel fast reactors. (author)

  8. Vibration analysis of reactor assembly internals for Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Vibration analysis of the reactor assembly components of 500 MWe Prototype Fast Breeder Reactor (PFBR) is presented. The vibration response of primary pump as well as dynamic forces developed at its supports are predicted numerically. The stiffness properties of hydrostatic bearing are determined by formulating and solving governing fluid and structural mechanics equations. The dynamic forces exerted by pump are used as input data for the dynamic response of reactor assembly components, mainly inner vessel, thermal baffle and control plug. Dynamic response of reactor assembly components is also predicted for the pressure fluctuations caused by sodium free level oscillations. Thermal baffle (weir shell) which is subjected to fluid forces developed at the associated sodium free levels is analysed by formulating and solving a set of non-linear equations for fluids, structures and fluid structure interaction (FSI). The control rod drive mechanism is analysed for response under flow induced forces on the parts subjected to cross flow in the zone just above the core top, taking into account FSI between sheaths of control and safety rod and absorber pin bundle. Based on the analysis results, it is concluded that the reactor assembly internals are free from any risk of mechanical as well as flow induced vibrations. (author)

  9. Study of some electrochemical properties of uranium in a molten fluoride medium. Application to the determination of the U(IV)/U(III) ratio in the fuel of a fused salt breeder reactor

    International Nuclear Information System (INIS)

    The aim of this work is to check the possibility of determining the U(IV)/U(III) ratio by electrochemical techniques in the fused LiF-BeF2-ThF4 mixture, solvent of a fused salt breeder reactor. The electrochemical properties of uranium systems were studied. The electrochemical reduction of U(IV) in LiF-BeF2-ThF4 at 615 0C was studied by linear and cyclic potential variation voltamperometry, constant current voltamperometry and pulse voltamperometry (normal and derived mode). The results obtained at a molybdenum electrode show two successive electron transfers. The formation of U-Mo alloy is observed. The oxidation of U(IV) was observed at a vitreous carbon electrode. The results show that the uranium V formed is unstable, their interpretation suggesting that U(IV) dismutes into U(IV) and probably into gaseous UF6. In view of these results it was possible to determine the U(IV)/U(III) ratio by the use of the above electrochemical techniques

  10. Status of national programmes on fast breeder reactors. Eighteenth annual meeting, Vienna, Austria, 16-19 April 1985

    International Nuclear Information System (INIS)

    The Eighteenth Annual Meeting on the Status of National Programmes in Member States of the IAEA on Fast Breeder Reactors had been held in April 1985. The representatives of the Member States and international organizations reported status and activities in the field of fast breeder reactors development and operation. A separate abstract was prepared for each of the 12 presentations of the meeting

  11. Cermet fuel reactors

    Energy Technology Data Exchange (ETDEWEB)

    Cowan, C.L.; Palmer, R.S.; Van Hoomissen, J.E.; Bhattacharyya, S.K.; Barner, J.O.

    1987-09-01

    Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. Key features of the cermet fueled reactor design are (1) the ability to achieve very high coolant exit temperatures, and (2) thermal shock resistance during rapid power changes, and (3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, thre is a potential for achieving a long operating life because of (1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and (2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core. In addition, the neutronic properties of the refractory materials assure that the reactor remains substantially subcritical under conditions of water immersion. It is concluded that cermet fueled reactors can be utilized to meet the power requirements for a broad range of advanced space applications. 4 refs., 4 figs., 3 tabs.

  12. Cermet fuel reactors

    International Nuclear Information System (INIS)

    Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. Key features of the cermet fueled reactor design are (1) the ability to achieve very high coolant exit temperatures, and (2) thermal shock resistance during rapid power changes, and (3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, thre is a potential for achieving a long operating life because of (1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and (2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core. In addition, the neutronic properties of the refractory materials assure that the reactor remains substantially subcritical under conditions of water immersion. It is concluded that cermet fueled reactors can be utilized to meet the power requirements for a broad range of advanced space applications. 4 refs., 4 figs., 3 tabs

  13. NUCLEAR REACTOR FUEL ELEMENT

    Science.gov (United States)

    Wheelock, C.W.; Baumeister, E.B.

    1961-09-01

    A reactor fuel element utilizing fissionable fuel materials in plate form is described. This fuel element consists of bundles of fuel-bearing plates. The bundles are stacked inside of a tube which forms the shell of the fuel element. The plates each have longitudinal fins running parallel to the direction of coolant flow, and interspersed among and parallel to the fins are ribs which position the plates relative to each other and to the fuel element shell. The plate bundles are held together by thin bands or wires. The ex tended surface increases the heat transfer capabilities of a fuel element by a factor of 3 or more over those of a simple flat plate.

  14. FUEL ASSAY REACTOR

    Science.gov (United States)

    Spinrad, B.I.; Sandmeier, H.A.; Martens, F.H.

    1962-12-25

    A reactor having maximum sensitivity to perturbations is described comprising a core consisting of a horizontally disposed, rectangular, annular fuel zone containing enriched uranium dioxide dispersed in graphite, the concentration of uranium dioxide increasing from the outside to the inside of the fuel zone, an internal reflector of graphite containing an axial test opening disposed within the fuel zone, an external graphite reflector, means for changing the neutron spectrum in the test opening, and means for measuring perturbations in the neutron flux caused by the introduction of different fuel elements into the test opening. (AEC)

  15. Characteristics of fission products behavior on a severe accident in fast breeder reactor

    International Nuclear Information System (INIS)

    Japan Nuclear Energy Safety Organization (JNES) has been developing the ACTOR code for the analysis of the fission products behavior under the severe accident condition to apply the probabilistic safety assessment to fast breeder reactor plants. Major analysis models of the ACTOR code were validated and adjusted by related experimental results. The fission products behavior on PLOHS (Protected Loss of Heat Sink) sequence which is one of the typical severe accidents in FBR plant was analyzed by using the ACTOR code. It was confirmed that the ACTOR had an enough capability to analyze the fission products behavior during severe accident. From the analysis results of PLOHS, it was confirmed that cesium is transferred to the cover gas region much more than iodine because iodine which is one of halogen connects to sodium easily and is retained in the coolant. Therefore, cesium is important and it is needed to examine the necessity to treat cesium as one of FPs considered in reactor establishment permission for FBR plant. Thus, cesium transfer behavior in sodium during the rare gas bubbles rise from fuel to the cover gas region was confirmed to be very important. And JNES started study including validation test about cesium transfer behavior with Hokkaido University. (author)

  16. Fast reactors using molten chloride salts as fuel

    International Nuclear Information System (INIS)

    This report deals with a rather exotic ''paper reactor'' in which the fuel is in the form of molten chlorides. (a) Fast breeder reactor with a mixed fuel cycle of thorium/uranium-233 and uranium 238/plutonium in which all of the plutonium can be burned in situ and in which a denatured mixture of uranium-233 and uranium-238 is used to supply further reactors. The breeding ratio is relatively high, 1.58 and the specific power is 0.75 GW(th)/m3 of core. (b) Fast breeder reactor with two and three zones (internal fertile zone, intermediate fuel zone, external fertile zone) with an extremely high breeding ratio of 1.75 and a specific power of 1.1 GW(th)/m3 of core. (c) Extremely high flux reactor for the transmutation of the fission products: strontium-90 and caesium-137. The efficiency of transmutation is approximately 15 times greater than the spontaneous beta decay. This high flux burner reactor is intended as part of a complex breeder/burner system. (d) Internally cooled fast breeder in which the cooling agent is the molten fertile material, the same as in the blanket zone. This reactor has a moderate breeding ratio of 1.38, a specific power of 0.22 GW(th)/m3 of core and very good inherent safety properties. All of these reactors have the fuel in the form of molten chlorides: PuCl3 as fissile, UCl3 as fertile (if needed) and NaCl as dilutent. The fertile material can be 238UCl3 as fertile and NaCl as dilutent. In mixed fuel cycles the 233UCl3 is also a fissile component with 232ThCl4 as the fertile constituent

  17. Final report for the Light Water Breeder Reactor proof-of-breeding analytical support project

    Energy Technology Data Exchange (ETDEWEB)

    Graczyk, D.G.; Hoh, J.C.; Martino, F.J.; Nelson, R.E.; Osudar, J.; Levitz, N.M.

    1987-05-01

    The technology of breeding /sup 233/U from /sup 232/Th in a light water reactor is being developed and evaluated by the Westinghouse Bettis Atomic Power Laboratory (BAPL) through operation and examination of the Shippingport Light Water Breeder Reactor (LWBR). Bettis is determining the end-of-life (EOL) inventory of fissile uranium in the LWBR core by nondestructive assay of a statistical sample comprising approximately 500 EOL fuel rods. This determination is being made with an irradiated-fuel assay gauge based on neutron interrogation and detection of delayed neutrons from each rod. The EOL fissile inventory will be compared with the beginning-of-life fissile loading of the LWBR to determine the extent of breeding. In support of the BAPL proof-of-breeding (POB) effort, Argonne National Laboratory (ANL) carried out destructive physical, chemical, and radiometric analyses on 17 EOL LWBR fuel rods that were previously assayed with the nondestructive gauge. The ANL work included measurements on the intact rods; shearing of the rods into pre-designated contiguous segments; separate dissolution of each of the more than 150 segments; and analysis of the dissolver solutions to determine each segment's uranium content, uranium isotopic composition, and loading of selected fission products. This report describes the facilities in which this work was carried out, details operations involved in processing each rod, and presents a comprehensive discussion of uncertainties associated with each result of the ANL measurements. Most operations were carried out remotely in shielded cells. Automated equipment and procedures, controlled by a computer system, provided error-free data acquisition and processing, as well as full replication of operations with each rod. Despite difficulties that arose during processing of a few rod segments, the ANL destructive-assay results satisfied the demanding needs of the parent LWBR-POB program.

  18. Final report for the Light Water Breeder Reactor proof-of-breeding analytical support project

    International Nuclear Information System (INIS)

    The technology of breeding 233U from 232Th in a light water reactor is being developed and evaluated by the Westinghouse Bettis Atomic Power Laboratory (BAPL) through operation and examination of the Shippingport Light Water Breeder Reactor (LWBR). Bettis is determining the end-of-life (EOL) inventory of fissile uranium in the LWBR core by nondestructive assay of a statistical sample comprising approximately 500 EOL fuel rods. This determination is being made with an irradiated-fuel assay gauge based on neutron interrogation and detection of delayed neutrons from each rod. The EOL fissile inventory will be compared with the beginning-of-life fissile loading of the LWBR to determine the extent of breeding. In support of the BAPL proof-of-breeding (POB) effort, Argonne National Laboratory (ANL) carried out destructive physical, chemical, and radiometric analyses on 17 EOL LWBR fuel rods that were previously assayed with the nondestructive gauge. The ANL work included measurements on the intact rods; shearing of the rods into pre-designated contiguous segments; separate dissolution of each of the more than 150 segments; and analysis of the dissolver solutions to determine each segment's uranium content, uranium isotopic composition, and loading of selected fission products. This report describes the facilities in which this work was carried out, details operations involved in processing each rod, and presents a comprehensive discussion of uncertainties associated with each result of the ANL measurements. Most operations were carried out remotely in shielded cells. Automated equipment and procedures, controlled by a computer system, provided error-free data acquisition and processing, as well as full replication of operations with each rod. Despite difficulties that arose during processing of a few rod segments, the ANL destructive-assay results satisfied the demanding needs of the parent LWBR-POB program

  19. Leakage limits for inflatable seals of sodium cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Highlights: • All possible types/modes of gas escape covered. • Limits include simultaneous contributions from bypass and permeation leakage modes. • Leakage of radioactive cover gas with fission products assumed. • Possibility of sodium frost deposition in sealed gap considered. • Cover gas activity decay during fuel handling and relative importance of types/modes of leakage considered for realistic results and simpler seal design. -- Abstract: Estimation and stipulation of allowable leakage for inflatable seals of 500 MWe Prototype Fast Breeder Reactor is depicted. Leakage limits are specified using a conservative approach, which assumes escape of radioactive cover gas with fission products across the seals in bypass and permeation modes and possibility of sodium frost deposition in sealed gaps because of permeation leakage of inflation gas. Procedures to arrive at the allowable leakages of argon cover gas (normal-operation/fuel-handling: 10−3/10−2 scc/s/m length of seal) and argon inflation gas (10−3 scc/s/m length of seal) is described

  20. Tritium-assisted fusion breeders

    International Nuclear Information System (INIS)

    This report undertakes a preliminary assessment of the prospects of tritium-assisted D-D fuel cycle fusion breeders. Two well documented fusion power reactor designs - the STARFIRE (D-T fuel cycle) and the WILDCAT (Cat-D fuel cycle) tokamaks - are converted into fusion breeders by replacing the fusion electric blankets with 233U producing fission suppressed blankets; changing the Cat-D fuel cycle mode of operation by one of the several tritium-assisted D-D-based modes of operation considered; adjusting the reactor power level; and modifying the resulting plant cost to account for the design changes. Three sources of tritium are considered for assisting the D-D fuel cycle: tritium produced in the blankets from lithium or from 3He and tritium produced in the client fission reactors. The D-D-based fusion breeders using tritium assistance are found to be the most promising economically, especially the Tritium Catalyzed Deuterium mode of operation in which the 3He exhausted from the plasma is converted, by neutron capture in the blanket, into tritium which is in turn fed back to the plasma. The number of fission reactors of equal thermal power supported by Tritium Catalyzed Deuterium fusion breeders is about 50% higher than that of D-T fusion breeders, and the profitability is found to be slightly lower than that of the D-T fusion breeders

  1. The fusion breeder

    International Nuclear Information System (INIS)

    The fusion breeder is a fusion reactor designed with special blankets to maximize the transmutation by 14 MeV neutrons of uranium-238 to plutonium or thorium to uranium-233 for use as a fuel for fission reactors. Breeding fissile fuels has not been a goal of the U.S. fusion energy program. This paper suggests it is time for a policy change to make the fusion breeder a goal of the U.S. fusion program and the U.S. nuclear energy program. There is wide agreement that many approaches will work and will produce fuel for five equal-sized LWRs, and some approach as many as 20 LWRs at electricity costs within 20% of those at today's price of uranium ($30/lb of U3O8). The blankets designed to suppress fissioning, called symbiotes, fusion fuel factories, or just fusion breeders, will have safety characteristics more like pure fusion reactors and will support as many as 15 equal power LWRs. The blankets designed to maximize fast fission of fertile material will have safety characteristics more like fission reactors and will support 5 LWRs. This author strongly recommends development of the fission suppressed blanket type, a point of view not agreed upon by everyone. There is, however, wide agreement that, to meet the market price for uranium which would result in LWR electricity within 20% of today's cost with either blanket type, fusion components can cost severalfold more than would be allowed for pure fusion to meet the goal of making electricity alone at 20% over today's fission costs. Also widely agreed is that the critical-pathitem for the fusion breeder is fusion development itself; however, development of fusion breeder specific items (blankets, fuel cycle) should be started now in order to have the fusion breeder by the time the rise in uranium prices forces other more costly choices

  2. Compilation of data and descriptions for United States and foreign liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    This document is a compilation of design and engineering information pertaining to liquid metal cooled fast breeder reactors which have operated, are operating, or are currently under construction, in the United States and abroad. All data has been taken from publicly available documents, journals, and books

  3. Recommendations concerning models and parameters best suited to breeder reactor environmental radiological assessments

    International Nuclear Information System (INIS)

    Recommendations are presented concerning the models and parameters best suited for assessing the impact of radionuclide releases to the environment by breeder reactor facilities. These recommendations are based on the model and parameter evaluations performed during this project to date. Seven different areas are covered in separate sections

  4. Engineering review of the core support structure of the Gas Cooled Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    None

    1978-09-01

    The review of the core support structure of the gas cooled fast breeder reactor (GCFR) covered such areas as the design criteria, the design and analysis of the concepts, the development plan, and the projected manufacturing costs. Recommendations are provided to establish a basis for future work on the GCFR core support structure.

  5. Atoms in Appalachia. Historical report on the Clinch River Breeder Reactor site

    Energy Technology Data Exchange (ETDEWEB)

    Schaffer, D

    1982-01-01

    The background information concerning the acquisition of the land for siting the Clinch River Breeder Reactor is presented. Historical information is also presented concerning the land acquisition for the Oak Ridge facilities known as the Manhattan Project during World War II.

  6. Recommendations concerning models and parameters best suited to breeder reactor environmental radiological assessments

    Energy Technology Data Exchange (ETDEWEB)

    Miller, C.W.; Baes, C.F. III; Dunning, D.E. Jr.

    1980-05-01

    Recommendations are presented concerning the models and parameters best suited for assessing the impact of radionuclide releases to the environment by breeder reactor facilities. These recommendations are based on the model and parameter evaluations performed during this project to date. Seven different areas are covered in separate sections.

  7. Fast reactors using molten chloride salts as fuel

    International Nuclear Information System (INIS)

    This report deals with a rather exotic 'paper reactor', in which the fuel is in the form of molten chlorides. (a) Fast breeder reactor with a mixed fuel cycle of thorium/uranium-233 and uranium 238/plutonium in which all of the plutonium can be burned in situ and in which a denatured mixture of uranium-233 and uranium-238 is used to supply further reactors. The breeding ratio is relatively high, 1.58 and the specific power is 0.75 GW(th)/m3 of core. (b) Fast breeder reactor with two and three zones (internal fertile zone, intermediate fuel zone, external fertile zone) with an extremely high breeding ratio of 1.75 and a specific power of 1.1 GW(th)/m3 of core. (c) Extremely high flux reactor for the transmutation of the fission products: strontium-90 and caesium-137. The efficiency of transmutation is approximately 15 times greater than the spontaneous beta decay. This high flux burner reactor is intended as part of a complex breeder/burner system. (d) Internally cooled fast breeder in which the cooling agent is the molten fertile material, the same as in the blanket zone. This reactor has a moderate breeding ratio of 1.38, a specific power of 0.22 GW(th)/m3 of core and very good inherent safety properties. All of these reactors have the fuel in the form of molten chlorides: PuCl3 as fissile, UCl3 as fertile (if needed) and NaCl as dilutent. (Auth.)

  8. Status of fast breeder reactor development in the Federal Republic of Germany, Belgium and the Netherlands

    International Nuclear Information System (INIS)

    The results of activities carried out by the DeBeNe partners in 1988 have been compiled in this report. The 1989 KNK II experimental program will focus on the management of fuel element failures. This includes in particular post-irradiation examinations in the hot cells and the THIBO experiments (thermally induced fuel rod oscillation). For this program, nine permits were received in 1988 for the installation and operation of test systems, including a new facility for localizing failed fuel elements. Increasing the purity of sodium is the purpose of a cesium trap recently installed, and of modifications to an oxygen probe operated for test purposes. The SNR 300 project is being continued within the framework of the ''holding phase''. The objective of this phase of operation is to keep the reactor plant in the loading and operational states and execute planning within the licensing procedure necessary to obtain the next partial permit. R and D work was concentrated on fuel and materials development, safety, physics, and components development. Materials for fuel, blanket, and absorber elements were studied and further developed with a view to very high burnup. In the sector of physics, the engineering design and the nuclear design of large breeders call for a reduction of the margins of uncertainty in reliable predictions of the major reactor parameters. The development of the joint European cell code, ECCO (European Cell Code) has progressed far enough in the meantime to allow the criticality parameter, keff, of a cell arrangement to be calculated with ECCO for the first time at UKAEA Winfrith, the central agency for carrying out the development work. One of the major aspects covered in safety studies is the complex of fuel rod failures, loss of coolant flow, and power transients as possible causes of accidents. Studies conducted into the management of credible accident consequences were concentrated, among other topics, on the behaviour of aerosols, sodium fires

  9. History of fast reactor fuel development

    International Nuclear Information System (INIS)

    The first fast breeder eactors, constructed in the 1945-1960 time period, used metallic fuels composed of uranium, plutonium, or their alloys. They were chosen because most existing reactor operating experience had been obtained on metallic fuels and because they provided the highest breeding ratios. Difficulties in obtaining adequate dimensional stability in metallic fuel elements under conditions of high fuel burnup led in the 1960s to the virtual worldwide choice of ceramic fuels. Although ceramic fuels provide lower breeding performance, this objective is no longer an important consideration in most national programs. Mixed uranium and plutonium dioxide became the ceramic fuel that has received the widest use. The more advanced ceramic fuels, mixed uranium and plutonium carbides and nitrides, continue under development. More recently, metal fuel elements of improved design have joined ceramic fuels in achieving goal burnups of 15 to 20 percent. Low-swelling fuel cladding alloys have also been continuously developed to deal with the unexpected problem of void formation in stainless steels subjected to fast neutron irradiation, a phenomenon first observed in the 1960s. (orig.)

  10. Engineering design and development for prototype fast breeder reactor (PFBR) shielding experiments at Apsara

    International Nuclear Information System (INIS)

    Prototype fast breeder reactor (PFBR) houses radial shields inside the reactor vessel which consists of many layers of steel and borated graphite within sodium coolant so as to reduce the neutron flux impingement on Intermediate Heat Exchanger (IHX) (also located inside the reactor vessel) to an acceptable limit. In order to cross check the uncertainties involved in theoretical shielding calculations and neutron cross-section data used, IGCAR proposed to carry out various shielding experiments at Apsara reactor to simulate the theoretical shielding configuration. The experiments would also provide bias factors for detailed shielding design calculations. The shielding experiments were planned to be carried out at Apsara shielding corner with reactor core brought to C-dash (C) position. The neutron flux intensity in the shielding corner was inadequate for the purpose of carrying out experiments. Hence the neutron flux level was enhanced to the order of 1010 n/cm2/s by replacing the water column between the core edge and SS liner of Apsara pool on the shielding corner side with an air filled leak tight aluminium box. The fuel loading in the reactor core was also modified to increase neutron flux intensity towards aluminium box. The neutron flux emerging out of the pool into the shielding corner is essentially a thermal neutron spectrum, which was converted into a typical fast reactor leakage neutron spectrum with the help of converter assemblies (CAs ). The converter assemblies were made of depleted uranium and the assemblies were installed on a CA trolley. The CA trolley was positioned outside Apsara pool in the shielding corner. The models of proposed shields manufactured from various shielding materials viz. sodium, steel, borated graphite and boron carbide were installed on a shield model (SM) trolley. The SM trolley was positioned behind CA trolley. Shield models had provisions for irradiating in any foils which were used for measuring the neutron attenuation

  11. Possible types of breeders with thorium cycle

    International Nuclear Information System (INIS)

    Neutronics calculations of simplified homogeneous reactor models show the possibility that metal-fueled LMFBRs and coated particle fueled gas cooled reactors achieve reactor doubling times of around 10 years with the thorium cycle. Three concepts of gas-cooled thorium cycle breeders are discused. (Author)

  12. Status of National Programmes on Fast Breeder Reactors. International Working Group on Fast Reactors Twenty-First Annual Meeting, Seattle, USA, 9-12 May 1988

    International Nuclear Information System (INIS)

    The following papers on the status of national programmes on fast breeder reactors are presented in this report: Fast breeder reactor development in France during 1987; Status of fast breeder reactor development in the Federal Republic of Germany, Belgium and the Netherlands; A review of the Indian fast reactor programme; A review of the Italian fast reactor programme; A review of the fast reactor programme in Japan; Status of fast reactor activities in the USSR; A review of the United Kingdom fast reactor programme; Status of liquid metal reactor development in the United States of America; Review of activities of the Commission of European Communities relating to fast reactors in 1987; European co-operation in the field of fast reactor research and development — 1987 progress report; A review of fast reactor activities in Switzerland

  13. Pebble Bed Reactor: core physics and fuel cycle analysis

    International Nuclear Information System (INIS)

    The Pebble Bed Reactor is a gas-cooled, graphite-moderated high-temperature reactor that is continuously fueled with small spherical fuel elements. The projected performance was studied over a broad range of reactor applicability. Calculations were done for a burner on a throwaway cycle, a converter with recycle, a prebreeder and breeder. The thorium fuel cycle was considered using low, medium (denatured), and highly enriched uranium. The base calculations were carried out for electrical energy generation in a 1200 MW/sub e/ plant. A steady-state, continuous-fueling model was developed and one- and two-dimensional calculations were used to characterize performance. Treating a single point in time effects considerable savings in computer time as opposed to following a long reactor history, permitting evaluation of reactor performance over a broad range of design parameters and operating modes

  14. Overview of pool hydraulic design of Indian prototype fast breeder reactor

    Indian Academy of Sciences (India)

    K Velusamy; P Chellapandi; S C Chetal; Baldev Raj

    2010-04-01

    Thermal hydraulics plays an important role in the design of liquid metal cooled fast breeder reactor components, where thermal loads are dominant. Detailed thermal hydraulic investigations of reactor components considering multi-physics heat transfer are essential for choosing optimum designs among the various possibilities. Pool hydraulics is multi-dimensional in nature and simple one-dimensional treatment for the same is often inadequate. Computational Fluid Dynamics (CFD) plays a critical role in the design of pool type reactors and becomes an increasingly popular tool, thanks to the advancements in computing technology. In this paper, thermal hydraulic characteristics of a fast breeder reactor, design limits and challenging thermal hydraulic investigations carried out towards successful design of Indian Prototype Fast Breeder Reactor (PFBR) that is under construction, are highlighted. Special attention is paid to phenomena like thermal stratification, thermal stripping, gas entrainment, inter-wrapper flow in decay heat removal and multiphysics cellular convection. The issues in these phenomena and the design solutions to address them satisfactorily are elaborated. Experiments performed for special phenomena, which are not amenable for CFD treatment and experiments carried out for validation of the computer codes have also been described.

  15. Engineering development studies for molten-salt breeder reactor processing No. 18

    International Nuclear Information System (INIS)

    A water--mercury system was used to study the effect of geometric variations on mass transfer rates in rectangular contractors similar to those proposed for the molten-salt breeder reactor (MSBR) fuel reprocessing scheme. Since mass transfer rates were not accurately predicted by the Lewis correlation, other correlations were investigated. A correlation which was found to fit the experimental results is given. Mass transfer rates are being measured in a fluoride salt--bismuth contactor. Experimental results indicate that the mass transfer rates in the salt--bismuth system fall between the Lewis correlation and the modified correlation given above. Autoresistance heating tests were continued in the fluorinator mock-up using LiF--BeF2--ThF4 (72-16-12 mole percent) salt. The equipment was returned to operating condition, and five experiments were run. Although correct steady-state operation was not achieved, the results were encouraging. A two-dimensional electrical analog was constructed to study current flow through the electrode sidearm and other critical areas of the test vessel. These studies indicate that no regions of abnormally high current density existed in the first nine runs with the present autoresistance heating equipment. Localized heating had previously been the suspected cause for the failure to achieve proper operation of this equipment. (U.S.)

  16. The long-term future for civilian nuclear power generation in France: The case for breeder reactors. Breeder reactors: The physical and physical chemistry parameters, associate material thermodynamics and mechanical engineering: Novelties and issues

    Science.gov (United States)

    Dautray, Robert

    2011-06-01

    The author firstly gives a summary overview of the knowledge base acquired since the first breeder reactors became operational in the 1950s. "Neutronics", thermal phenomena, reactor core cooling, various coolants used and envisioned for this function, fuel fabrication from separated materials, main equipment (pumps, valves, taps, waste cock, safety circuits, heat exchange units, etc.) have now attained maturity, sufficient to implement sodium cooling circuits. Notwithstanding, the use of metallic sodium still raises certain severe questions in terms of safe handling (i.e. inflammability) and other important security considerations. The structural components, both inside the reactor core and outside (i.e. heat exchange devices) are undergoing in-depth research so as to last longer. The fuel cycle, notably the refabrication of fuel elements and fertile elements, the case of transuranic elements, etc., call for studies into radiation induced phenomena, chemistry separation, separate or otherwise treatments for materials that have different radioactive, physical, thermodynamical, chemical and biological properties. The concerns that surround the definitive disposal of certain radioactive wastes could be qualitatively improved with respect to the pressurized water reactors (PWRs) in service today. Lastly, the author notes that breeder reactors eliminate the need for an isotope separation facility, and this constitutes a significant contribution to contain nuclear proliferation. Among the priorities for a fully operational system (power station - the fuel cycle - operation-maintenance - the spent fuel pool and its cooling system-emergency cooling system-emergency electric power-transportation movements-equipment handling - final disposal of radioactive matter, independent safety barriers), the author includes materials (fabrication of targets, an irradiation and inspection instrument), the chemistry of all sorting processes, equipment "refabrication" or rehabilitation

  17. Fast breeder reactors: The state of materials subjected to high energy radiation, high local pressure and temperature, gradients and their mechanical properties adapted to the resultant constraints

    International Nuclear Information System (INIS)

    The motivations to realize nuclear breeder reactors are developed in the present context of a strong growth in electronuclear power stations in many countries, using mostly moderated and water cooled reactors. The past studies can be of a substantial profit in France and, to a lesser degree, in other countries of the European Union. However, to use fully the 238 uranium isotope, the materials for these breeders must withstand much harder radiation than those for water reactors. The power densities and thermal gradients will also be much more intense. The mechanical stresses, both static and dynamic, will be large and will act on materials with altered mechanical properties. Fuel elements will have to be produced with materials already irradiated several times and therefore showing such alterations. A field of studies concerning materials and their mechanical behavior in new and severe conditions is sketched here, both in construction and working conditions, together with proposed necessary instrumentation and research orientations. (authors)

  18. Supplement to Final Environmental Statement related to construction and operation of Clinch River Breeder Reactor Plant, Docket No. 50-537

    International Nuclear Information System (INIS)

    In February 1977, the Office of Nuclear Reactor Regulation issued a Final Environmental Statement (FES) (NUREG-0139) related to the construction and operation of the proposed Clinch River Breeder Reactor Plant (CRBRP). Since the FES was issued, additional data relative to the site and its environs have been collected, several modifications have been made to the CRBRP design, and its fuel cycle, and the timing of the plant construction and operation has been affected in accordance with deferments under the DOE Liquid Metal Fast Breeder Reactor (LMFBR) program. These changes are summarized and their environmental significance is assessed in this document. The reader should note that this document generally does not repeat the substantial amount of information in the FES which is still current; hence, the FES should be consulted for a comprehensive understanding of the staff's environmental review of the CRBRP project

  19. k-eff of the Bn-350 reactor fuel by transportation

    International Nuclear Information System (INIS)

    There is packaging of nuclear fuel on the BN-350 fast breeder reactor, Actau, now. The analysis of criticality while this procedure was done in the Safety Analysis Report . Keeping in mind the planning displacement of the fuel to a site of long-term storage, the criticality assessment of the fuel packed into transportation cask carried out in this paper

  20. Gaseous fuel reactor research

    Science.gov (United States)

    Thom, K.; Schneider, R. T.

    1977-01-01

    The paper reviews studies dealing with the concept of a gaseous fuel reactor and describes the structure and plans of the current NASA research program of experiments on uranium hexafluoride systems and uranium plasma systems. Results of research into the basic properties of uranium plasmas and fissioning gases are reported. The nuclear pumped laser is described, and the main results of experiments with these devices are summarized.

  1. Control rod calibration methods for fast breeder reactors applied to Phenix

    International Nuclear Information System (INIS)

    The control and the emergency shutdown of a fast breeder reactor depends essentially on control rods. For this reason, it is imperative to know exactly how much anti reactivity is introduced with the rods in the reactor core. Different methods have been compared in order to see if they are compatible with Phenix reactor. Their limits have been studied. The shadow and anti shadow effects that can the rods make one to the other and then their effective weight of the rods screen have been clarified. (N.C.)

  2. Preliminary physics design of accelerator-driven thorium cycle fast breeder reactor

    International Nuclear Information System (INIS)

    A preliminary reactor physics design of a lead cooled fast accelerator-driven system has been explored as a thorium-uranium cycle breeder reactor. The sub-critical reactor core operates at an effective neutron multiplication factor of 0.95 and when driven by 1 GeV proton beams of intensity 30 mA, produces about ∼ 900 MWth power. Variation of total thermal power, 233U inventory, Keff, radial and axial power distribution through the operating cycle as well as breeding ratio and doubling time are presented. (author)

  3. Methodical study of cost-benefit analyses of the liquid metal fast breeder reactor

    International Nuclear Information System (INIS)

    Six American cost-benefit analyses (CBA) of nuclear energy and, in particular, of the Liquid Metal Fast Breeder Reactor (LMFBR) were analysed under the aspect of their methodical difficulties. Two different methodical approaches can be discerned which are related to two completely different applications, according to which the advantages and disadvantages of the breeder reactor are estimated in line with the basic concept of cost-benefit analysis. The analytical methods used to justify the continuation of the breeder-related research programme reveal that the specific energy-related technological and economic conditions of the geographic region considered have to be taken into account. The results of a CBA performed for the USA can therefore not be transferred to the Federal Republic of Germany. Due to the in part strongly differing quantitative results the analyses reviewed do not suggest a clear and final decision in favour of the continuation of the American LMFBR research programme to the extent envisaged. In addition, neither by a positive nor by a negative overall result of the analysis can it be concluded that no other advanced electricity generating technology would have a more favourable cost-benefit ratio, or that the breeder-related research activities, which have been pursued for several years already, should be discontinued. (orig.)

  4. Laboratory analysis of sodium and related materials in the United States Breeder Reactor Program

    Energy Technology Data Exchange (ETDEWEB)

    Keough, R.F.; Ebersole, E.R.

    1978-04-01

    It is essential for the long term operation of a Liquid Metal Fast Breeder Reactor (LMFBR) to control the levels of some sodium and cover gas impurities and to monitor the levels of many others. Chemical analyses are the means of confirming that specific impurities are within established limits. Changes in impurity levels can be most useful in evaluating purification systems and detecting problems such as leaks or excessive corrosion. Extensive analyses are also required with experimental sodium systems related to the Breeder Reactor Program (BRP) since reactor quality sodium is needed to insure relevancy of test data. All of the laboratories use the same procedures for sodium and cover gas analysis. Each year the representatives of the eight laboratories meet to discuss problems they may have and to suggest improvements they would like to see in the methods. Many of these methods are described and their application related to the operation of the Experimental Breeder Reactor II (EBR-II) and Fast Flux Test Facility (FFTF).

  5. Build-up and decay of fuel actinides in the fuel cycle of nuclear reactors

    International Nuclear Information System (INIS)

    For boiling water reactors, pressurized light-water reactors, pressure-tube-type heavy water reactors, high-temperature gas-cooled reactors, and sodium-cooled fast breeder reactors, uranium fueled and mixed-oxide fueled, each of 1000 MWe, the following have been studied: (1) quantities of plutonium and other fuel actinides built up in the reactor, (2) cooling behaviors of activities of plutonium and other fuel actinides in the spent fuels, and (3) activities of plutonium and other fuel actinides in the high-level reprocessing wastes as a function of storage time. The neutron cross section and decay data of respective actinide nuclides are presented, with their evaluations. For effective utilization of the uranium resources and easy reprocessing and high-level waste management, a thermal reactor must be fueled with uranium; the plutonium produced in a thermal reactor should be used in a fast reactor; and the plutonium produced in the blanket of a fast reactor is more appropriate for a fast reactor than that from a thermal reactor. (auth.)

  6. Fast breeder reactor reference system classification for the ENEA data bank

    International Nuclear Information System (INIS)

    This report contains the Reference System Classification (RSC) of fast breeder reactors: it provides a functional system breakdown of the reactor. For each system the following important characteristics are reported: the main function, the mode of operation, its location in the reactor, the main interface system, its main components and the component working environment (fluid and/or atmosphere type). The RSC represent a basic step in organizing the ENEA data bank for the registration and processing of reliability data on typical fast reactor components; it provides a functional component breakdown and represent a plant-unique identification in the process of omogenization of event-data coming from different reactors. In this report it was tried to take into account different generations of nuclear power plants, different plant layouts and solutions: in particular loop and pool reactors are separately treated

  7. Liquid-metal fast breeder reactor structural materials design considerations

    International Nuclear Information System (INIS)

    This paper gives a brief overview of the LMFBR, to describe its key components, addresses two key structural problems, reviews high-temperature materials utilized, and places bounds on expected operating conditions. The current status of materials utilization in the LMFBR is summarized as follows: with the exception of the reactor upper internals, design needs for the LMFBR can be met with currently approved Code materials; Inconel 718 can potentially solve the thermal striping problems in the reactor upper internals; temperature, stress-strain levels, and design lifetime of the LMFBR push currently approved Code materials toward their limits of usefulness

  8. Tritium solubility and permeation in high retention fusion reactor breeder elements

    International Nuclear Information System (INIS)

    As an alternative to the current philosophy of reducing the tritium inventory to a minimum by continuously extracting tritium from the breeder of a fusion reactor, an alternative design philosophy is examined in which tritium is contained within high retention breeder elements which can remain in the reactor for a substantial time. To prevent tritium diffusion through the clad of the element it is necessary to maintain a low tritium pressure within the element. Pressures of between 104 Pa and 1 Pa appear possible with an element containing a high solubility material provided it is kept below about 4000C. This should lead to a leakage into the coolant of between 10 Ci day-1 and 104 Ci day-1 which is considerably less than the 107 Ci day-1 in present designs. (author)

  9. Liquid-metal pumps for large-scale breeder-reactor plant (prototype pump)

    Energy Technology Data Exchange (ETDEWEB)

    Lindsay, M. (comp.)

    1976-07-01

    This report presents the recommended pump design for use in Large Scale Liquid Metal Fast Breeder Reactor plants. The base design for the pump will circulate 127,000 GPM of liquid sodium at temperatures up to 850/sup 0/F and with a total discharge head at the design point of 500 feet Na with an impeller that is 40 feet below the sodium seal. The pump design is predicated on developing an impeller design which will have a suction specific speed (S/sub n/) of about 20,000 with 20 feet NPSH available, which will result in a pump speed of 530 RPM at design conditions. The design is based on the technology developed in the design and fabrication of FFTF pumps, the design efforts for the Clinch River Breeder Reactor Pump design study and other technology.

  10. Tube sheet structural analysis of intermediate heat exchanger for fast breeder reactor 'Monju'

    International Nuclear Information System (INIS)

    The Prototype Fast Breeder Reactor 'Monju' is the first power generating fast breeder reactor in Japan. We have been designing the components of the plant for manufacturing. Among these is the intermediate heat exchanger (IHX) which exchanges heat between primary and secondary sodium loop. The tube sheet of IHX (shell to ligament junction) is a difficult area from the view point of structural strength design under elevated temperature. To validate the structural integrity of tube sheet we performed the series of inelastic analysis and tube sheet thermal shock test using test pieces and half scale model of actual design. The results of inelastic analyses showed there is little progressive deformation around shell to ligament structural discontinuous junction. Furthermore, thermal shock tests showed no increase of an accumulative deformation. By these analyses and experiments, structural reliability of tube sheet could be shown. (author)

  11. Internal fluid flow management analysis for Clinch River Breeder Reactor Plant sodium pumps

    International Nuclear Information System (INIS)

    The Clinch River Breeder Reactor Plant (CRBRP) sodium pumps are currently being designed and the prototype unit is being fabricated. In the design of these large-scale pumps for elevated temperature Liquid Metal Fast Breeder Reactor (LMFBR) service, one major design consideration is the response of the critical parts to severe thermal transients. A detailed internal fluid flow distribution analysis has been performed using a computer code HAFMAT, which solves a network of fluid flow paths. The results of the analytical approach are then compared to the test data obtained on a half-scale pump model which was tested in water. The details are presented of pump internal hydraulic analysis, and test and evaluation of the half-scale model test results

  12. Gel-sphere-pac reactor fuel fabrication and its application to a variety of fuels

    International Nuclear Information System (INIS)

    The gel-sphere-pac fuel fabrication option was evaluated for its possible application to commercial scale fuel fabrication for 19 fuel element designs that use oxide fuel in metal clad rods. The dry gel spheres are prepared at the reprocessing plant and are then calcined, sintered, inspected, and loaded into fuel rods and packed by low-energy vibration. A fuel smear density of 83 to 88% theoretical can be obtained. All fuel fabrication process steps were defined and evaluated from fuel receiving to finished fuel element shipping. The evaluation also covers the feasibility of the process, the current status of technology, estimates of the required time and cost to develop the technology to commercial status, and the safety and licensability of commercial scale plants. The primary evaluation was for a Light-Water Reactor fuel element containing (U,Pu)O2 fuel. The other 18 fuel element types - 3 for Light-Water Reactors, 1 for a Heavy-Water Reactor, 1 for a Gas-Cooled Fast Reactor, 7 for Liquid-Metal-Cooled Fast Breeder Reactors, and 3 pairs for Light-Water Prebreeder and Breeder Reactors - were compared with the Light-Water Reactor. The gel-sphere-pac option was found applicable to 17 of the 19 element types; the characteristics of a commercial scale plant were defined for these for making cost estimates for such plants. The evaluation clearly shows the gel-sphere-pac process to be a viable fuel fabrication option. Estimates indicate a significant potential fabrication cost advantage for the gel-sphere-pac process if a remotely operated and remotely maintained fuel fabrication plant is required

  13. System for fuel rod removal from a reactor module

    International Nuclear Information System (INIS)

    This patent describes a robotic system for remote underwater withdrawal of the fuel rods from fuel modules of a light water breeder reactor includes a collet/grapple assembly for gripping and removing fuel rods in each module, which is positioned by use of a winch and a radial support means attached to a vertical support tube which is mounted over the fuel module. A programmable logic controller in conjunction with a microcomputer, provides control for the accurate positioning and pulling force of the rod grapple assembly. Closed circuit television cameras are provided which aid in operator interface with the robotic system

  14. Status of the fast breeder reactor technology in China

    International Nuclear Information System (INIS)

    According to the Chinese long-term energy strategy the FBR development is strongly supported. In the near term nuclear programme it is intended to build the experimental First Fast Reactor (FFR) in the year 2000. Design work is in progress. (author). 1 ref., 6 figs, 8 tabs

  15. Application of microprocessor based controller in the Breeder Reactor Program

    International Nuclear Information System (INIS)

    This paper treats Argonne National Laboratory's experience using microprocessor based controllers presently in use on several control loops within the EBR-II reactor facility as well as tests being performed by these controllers. Also included is a discussion of the expandability, modularity, range of capabilities and higher level functions possible using such equipment

  16. Seismic parametric studies in a large scale prototype breeder reactor plant

    International Nuclear Information System (INIS)

    Seismic parametric studies were conducted for a large scale prototype breeder reactor plant (135C MW). The effects of plant configuration, soil stiffness and deep embedment were evaluated. Nuclear island interconnected structures on a common foundation mat with a symmetrical arrangement resulted in lower seismic responses. All other conditions being equal, soft sites are preferable to stiff sites. Deep embedment of the nuclear island could, in certain sites, result in a reduction of seismic responses. (orig.)

  17. Accident consequence studies for large fast breeder reactor containments built of concrete or steel

    International Nuclear Information System (INIS)

    A numerical analysis of accident consequences in a fast breeder reactor of commercial size after complete loss-of-heat-sink was performed, using the CONTAIN code. Two containment types were studied, which differ in the material used for shielding, support and confinement structures. It was found that the replacement of concrete as principal construction material by steel offers a significant potential for consequence mitigation in terms of thermal and pressure loads and of retention capability

  18. Present day design challenges exemplified by the Clinch River Breeder Reactor Plant

    International Nuclear Information System (INIS)

    The present day design challenges faced by the Clinch River Breeder Reactor Plant engineer result from two causes. The first cause is aspiration to achieve a design that will operate at conditions which are desirable for future LMFBRs in order for them to achieve low power costs and good breeding. The second cause is the licensing impact. Although licensing the CRBRP won't eliminate future licensing effort, many licensing questions will have been resolved and precedents set for the future LMFBR industry

  19. Application of mass-predictions to isotope-abundances in breeder-reactor cores

    International Nuclear Information System (INIS)

    The decay-heat and isotope composition of breeder reactor-cores is calculated at normal shut-down, and a core disintegration event. Using the ORIGRN-code, the influence of the most neutron-rich fission-yield nuclei is studied. Their abundances depend on the assumption about the nuclear data (mass and half-lives). The total decay-heat is not changed from any technically view-point. (orig.)

  20. Thermal insulation system design and fabrication specification (nuclear) for the Clinch River Breeder Reactor plant

    International Nuclear Information System (INIS)

    This specification defines the design, analysis, fabrication, testing, shipping, and quality requirements of the Insulation System for the Clinch River Breeder Reactor Plant (CRBRP), near Oak Ridge, Tennessee. The Insulation System includes all supports, convection barriers, jacketing, insulation, penetrations, fasteners, or other insulation support material or devices required to insulate the piping and equipment cryogenic and other special applications excluded. Site storage, handling and installation of the Insulation System are under the cognizance of the Purchaser

  1. Application of mass-predictions to isotope-abundances in breeder-reactor cores

    CERN Document Server

    Kirchner, G

    1981-01-01

    The decay-heat and isotope composition of breeder reactor-cores is calculated at normal shut-down, and a core disintegration event. Using the ORIGEN-code, the influence of the most neutron-rich fission-yield nuclei is studied. Their abundances depend on the assumption about the nuclear data (mass and half-lives). The total decay-heat is not changed from any technical viewpoint. (15 refs).

  2. Research and developments on nondestructive testing in fabrications of fast breeder reactor structural components in Japan

    International Nuclear Information System (INIS)

    Research and developments (R and D) have been conducted on the nondestructive testing techniques necessary for the construction of fast breeder reactor (FBR). Radiographic tests have been made on tube-tube plate welds and small-diameter tube welds, etc. Ultrasonic tests have been conducted on austenitic stainless steel welds. In the penetrant tests and magnetic particle tests, the investigations have been performed on the effects of various test factors on the test results

  3. Fast breeder reactor. The past, the present and the future. (7) History of fast reactor development in Japan - 2

    International Nuclear Information System (INIS)

    History and present state of fast breeder reactor was reviewed in series. As a history of fast reactor development in Japan - 2, this seventh lecture presented the development of the prototype FBR (MONJU) and design studies of the demonstration reactor. The MONJU started operation in 1994, but a sodium leakage in its secondary heat transfer system occurred during performance tests in 1995. It has not operated since and activities for restart are conducted. Since 1997 design studies of the demonstration FBR have been conducted to reflect the MONJU sodium leakage accident and also establish its economic competitiveness with advanced LWR. (T. Tanaka)

  4. Status of national programmes on fast breeder reactors. Nineteenth annual meeting, Kalpakkam, India, 11-14 March 1986

    International Nuclear Information System (INIS)

    The Nineteenth Annual Meeting on the Status of National Programmes in Member States of the IAEA on Fast Breeder Reactors had been held in March 1986. The representatives of the Member States and international organizations reported status and activities in the field of fast breeder reactors development and operation. A report on uranium supply and demand was also presented by the NEA/OECD. A separate abstract was prepared for each of the 11 presentations of the meeting

  5. Choice of rotatable plug seals for prototype fast breeder reactor: Review of historical perspectives

    International Nuclear Information System (INIS)

    Highlights: • Choice and arrangement of elastomeric inflatable and backup seals as primary and secondary barriers. • With survey (mid-1930s onwards) of reactor, sealing, R&D and rubber technology. • Load, reliability, safety, life and economy of seals and reactors are key factors. • PFBR blends concepts and experience of MOX fuelled FBRs with original solutions. • R&D indicates inflatable seal advanced fluoroelastomer pivotal in unifying nuclear sealing. - Abstract: Choice and arrangement of elastomeric primary inflatable and secondary backup seals for the rotatable plugs (RPs) of 500 MW (e), sodium cooled, pool type, 2-loop, mixed oxide (MOX) fuelled Prototype Fast Breeder Reactor (PFBR) is depicted with review of various historical perspectives. Static and dynamic operation, largest diameters (PFBR: ∼6.4 m, ∼4.2 m), widest gaps and variations (5 ± 2 mm) and demanding operating requirements make RP openings on top shield (TS) the most difficult to seal which necessitated extensive development from 1950s to early 1990s. Liquid metal freeze seals with life equivalent to reactor prevailed as primary barrier (France, Japan, U.S.S.R.) during pre-1980s in spite of bulk, cost and complexity due to the abilities to meet zero leakage and resist core disruptive accident (CDA). Redefinition of CDA as beyond design basis accident, tolerable leakage and enhanced economisation drive during post-1980s established elastomeric inflatable seal as primary barrier excepting in U.S.S.R. (MOX fuel, freeze seal) and U.S.A. (metallic fuel). Choice of inflatable seal for PFBR RPs considers these perspectives, inherent advantages of elastomers and those of inflatable seals which maximise seal life. Choice of elastomeric backup seal as secondary barrier was governed by reliability and minimisation as well as distribution of load (temperature, radiation, mist) to maximise seal life. The compact sealing combination brings the hanging RPs at about the same elevation to reduce

  6. Choice of rotatable plug seals for prototype fast breeder reactor: Review of historical perspectives

    Energy Technology Data Exchange (ETDEWEB)

    Sinha, N.K., E-mail: nksinha@igcar.gov.in; Raj, Baldev, E-mail: baldev.dr@gmail.com

    2015-09-15

    Highlights: • Choice and arrangement of elastomeric inflatable and backup seals as primary and secondary barriers. • With survey (mid-1930s onwards) of reactor, sealing, R&D and rubber technology. • Load, reliability, safety, life and economy of seals and reactors are key factors. • PFBR blends concepts and experience of MOX fuelled FBRs with original solutions. • R&D indicates inflatable seal advanced fluoroelastomer pivotal in unifying nuclear sealing. - Abstract: Choice and arrangement of elastomeric primary inflatable and secondary backup seals for the rotatable plugs (RPs) of 500 MW (e), sodium cooled, pool type, 2-loop, mixed oxide (MOX) fuelled Prototype Fast Breeder Reactor (PFBR) is depicted with review of various historical perspectives. Static and dynamic operation, largest diameters (PFBR: ∼6.4 m, ∼4.2 m), widest gaps and variations (5 ± 2 mm) and demanding operating requirements make RP openings on top shield (TS) the most difficult to seal which necessitated extensive development from 1950s to early 1990s. Liquid metal freeze seals with life equivalent to reactor prevailed as primary barrier (France, Japan, U.S.S.R.) during pre-1980s in spite of bulk, cost and complexity due to the abilities to meet zero leakage and resist core disruptive accident (CDA). Redefinition of CDA as beyond design basis accident, tolerable leakage and enhanced economisation drive during post-1980s established elastomeric inflatable seal as primary barrier excepting in U.S.S.R. (MOX fuel, freeze seal) and U.S.A. (metallic fuel). Choice of inflatable seal for PFBR RPs considers these perspectives, inherent advantages of elastomers and those of inflatable seals which maximise seal life. Choice of elastomeric backup seal as secondary barrier was governed by reliability and minimisation as well as distribution of load (temperature, radiation, mist) to maximise seal life. The compact sealing combination brings the hanging RPs at about the same elevation to reduce

  7. Assessment of a core meltdown in the gas-cooled fast breeder reactor with an upflow core

    International Nuclear Information System (INIS)

    This paper discusses the chronological sequence of events and supporting analysis of a postulated total loss of all coolant circulation in the gas-cooled fast breeder reactor (GCFR) with an upflow core. Redundant and diverse cooling systems are provided for decay heat removal, including pressurized natural circulation in the core auxiliary cooling system, which reduce the probability of this postulated event below the range of plant design bases. Nevertheless, this postulated accident has been considered so that the potential for consequence mitigation and containment margin could be investigated. Two distinct phases of the sequence are discussed: (1) the core response to a total loss of forced and natural coolant circulation and (2) the capability of the prestressed concrete reactor vessel (PCRV) to retain molten fuel debris. Specific design features of the GCFR which prevent recriticality and fuel vaporization due to fuel slumping are under investigation. Analytical work has been initiated to determine the potential for consequence mitigation in the PCRV and the containment. Several concepts for postaccident fuel containment have been identified and appear technically feasible

  8. Contained fissionly vaporized imploded fission explosive breeder reactor

    International Nuclear Information System (INIS)

    Disclosed is a nuclear reactor system which produces useful thermal power and breeds fissile isotopes wherein large spherical complex slugs containing fissile and fertile isotopes as well as vaporizing and tamping materials are exploded seriatim in a large containing chamber having walls protected from the effects of the explosion by about two thousand tons of slurry of fissile and fertile isotopes in molten alkali metal. The slug which is slightly sub-critical prior to its entry into the centroid portion of the chamber, then becomes slightly more than prompt-critical because of the near proximity of neutron-reflecting atoms and of fissioning atoms within the slurry. The slurry is heated by explosion of the slugs and serves as a working fluid for extraction of heat energy from the reactor. Explosive debris is precipitated from the slurry and used for the fabrication of new slugs

  9. Nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    The description is given of a nuclear reactor fuel assembly comprising fuel elements arranged in a supporting frame composed of two end pieces, one at the top and the other at the bottom, on which are secured the ends of a number of vertical tubes, each end piece comprising a plane bottom on which two series of holes are made for holding the tubes and for the passage of the coolant. According to the invention, the bottom of each end piece is fixed to an internal plate fitted with the same series of holes for holding the tubes and for the fluid to pass through. These holes are of oblong section and are fitted with fixing elements cooperating with corresponding elements for securing these tubes by transversal movement of the inside plate

  10. A decade of progress in fast reactor fuel

    International Nuclear Information System (INIS)

    This paper discusses the enormous strides that have been made in the World's knowledge of the behavior of fast reactor fuels and materials over the past 10 yr. With ∼ 12 sodium-cooled fast reactors operating in various countries and with the willingness of the international community to freely share the knowledge gained, the technological achievements have been staggering. There is no doubt that fast breeders can now be built and operated safely and effectively. Mixed-oxide fuel has been demonstrated to reliably meet its design objectives, and the breeder communities are now shifting emphasis to economic and safety improvements. In the US, metal fuel is being explored as a possible successor to mixed oxide, because metal fuel offers the potential for in-house reprocessing. The European Community, Japan, and the Soviet Union are all pursuing nitride fuel. Continued testing under fully prototypical conditions in a high-quality research reactor such as the FFTF is the key to the optimal selection of a fast reactor fuel system for meeting the energy needs of future generations

  11. Status of fast breeder reactor development in India

    International Nuclear Information System (INIS)

    The energy scenario and economic conditions in India are presented. India needs considerable energy for its rapid industrialisation with the liberal economic policy. Nuclear energy with FBR is the only large scale energy resource other than coal, available in the country. The present economic constraints have delayed the construction of new NPPs. The performance of operating reactors has improved considerably during the year. Operating experience of FBTR has been detailed particularly the reactivity incident and its investigations. Updated design of 500 MWe PFBR is presented. Various R and D works in support of FBR in the engineering, metallurgy, chemistry, reprocessing, safety etc. are detailed. (author)

  12. Seismic analysis of liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    This report is a general survey of the recent methods to predict the seismic structural behaviour of LMFBRs. It shall put into evidence the impact of seismic analysis on the design of the different structures of the reactor. This report is addressed to specialists and institutions of governmental organizations in industrialized and developing countries responsible for the design and operation of LMFBRs. The information presented should enable specialists in the R and D institutions and industries likely to be involved, to establish the correct course of the design and operation of LMFBRs. Also, the safety aspect of seismic risk are emphasized in the report. Refs and figs

  13. Fast breeder reactor program. Hearings before the Joint Economic Committee, Congress of the United States, Ninety-Fourth Congress, First Session

    International Nuclear Information System (INIS)

    The economics of the liquid metal fast breeder reactor (LMFBR) was the subject of hearings of the Joint Economic Committee, chaired by Sen. Hubert Humphrey. FY '76 funding for the breeder program is $450 million, the largest single item of the Federal energy program. Elmer B. Staats, U.S. Comptroller General, testified on the rising costs of demonstration facilities and pointed out that Federal agencies are required to make all estimates of costs and benefits in constant dollars rather than projecting for inflation. Staats recommended a joint ERDA-Congressional study of the possible use of foreign breeder technology. Sheldon Meyers of the Environmental Protection Agency, while not opposing the breeder program, recommended a delay to resolve three problem areas: (1) base energy demand projections; (2) timing of the commercial introduction of the LMFBR; and (3) uncertainties over possible benefits from the LMFBR program. Theodore B. Taylor, International Research and Technology Corp., discussed the costs and security safeguards of the LMFBR, which produces more spent fuel than the light water reactor. Other witnesses included Robert Seamans and officials from ERDA, Ralph Nader, and speakers from private study groups

  14. Possible types of breeders with thorium cycle

    International Nuclear Information System (INIS)

    Neutronics calculations of simplified homogeneous reactor models show the possibility that metal-fueled LMFBRs and coated particle fueled gas cooled reactors achieve doubling times of around 10 years with the thorium cycle. Three concepts of gas-cooled thorium cycle breeders are discussed. (Author)

  15. Civilian nuclear power on the drawing board: the development of Experimental Breeder Reactor-II

    International Nuclear Information System (INIS)

    On September 28, 2001 a symposium was held at Argonne National Laboratory as part of the festivities to mark the 100th birthday of Enrico Fermi. The symposium celebrated Fermi's ''contribution to the development of nuclear power'' and focused on one particular ''line of development'' resulting from Fermi's interest in power reactors: Argonne's fast reactor program. Symposium participants made many references to the ways in which the program was linked to Fermi, who led the team which created the world's first self-sustaining nuclear chain reaction. For example, one presentation featured an April, 1944 memo that described a meeting attended by Fermi and others. The memo came from the time when research on plutonium and the nuclear chain reaction at Chicago's WWII Metallurgical Laboratory was nearing its end. Even as other parts of the Manhattan Engineering Project were building on this effort to create the bombs that would end the war, Fermi and his colleagues were taking the first steps to plan the use of nuclear energy in the postwar era. After noting that Fermi ''viewed the use of [nuclear] power for the heating of cities with sympathy,'' the group outlined several power reactor designs. In the course of discussion, Fermi and his colleagues took the first steps in conjuring the vision that would later be brought to life with Experimental Breeder Reactor I (EBR-I) and Experimental Breeder Reactor II (EBR-II), the celebrated achievements of the Argonne fast reactor program. Group members considered various schemes for a breeder reactor in which the relatively abundant U-238 would be placed near a core of fissionable material. The reactor would be a fast reactor; that is, neutrons would not be moderated, as were most wartime reactors. Thus, the large number of neutrons emitted in fast neutron fission would hit the U-238 and create ''extra'' fissionable material, that is, more than ''invested,'' and at the same time produce power. The group identified the problem of

  16. Civilian nuclear power on the drawing board: the development of Experimental Breeder Reactor-II.

    Energy Technology Data Exchange (ETDEWEB)

    Westfall, C.

    2003-02-20

    On September 28, 2001 a symposium was held at Argonne National Laboratory as part of the festivities to mark the 100th birthday of Enrico Fermi. The symposium celebrated Fermi's ''contribution to the development of nuclear power'' and focused on one particular ''line of development'' resulting from Fermi's interest in power reactors: Argonne's fast reactor program. Symposium participants made many references to the ways in which the program was linked to Fermi, who led the team which created the world's first self-sustaining nuclear chain reaction. For example, one presentation featured an April, 1944 memo that described a meeting attended by Fermi and others. The memo came from the time when research on plutonium and the nuclear chain reaction at Chicago's WWII Metallurgical Laboratory was nearing its end. Even as other parts of the Manhattan Engineering Project were building on this effort to create the bombs that would end the war, Fermi and his colleagues were taking the first steps to plan the use of nuclear energy in the postwar era. After noting that Fermi ''viewed the use of [nuclear] power for the heating of cities with sympathy,'' the group outlined several power reactor designs. In the course of discussion, Fermi and his colleagues took the first steps in conjuring the vision that would later be brought to life with Experimental Breeder Reactor I (EBR-I) and Experimental Breeder Reactor II (EBR-II), the celebrated achievements of the Argonne fast reactor program. Group members considered various schemes for a breeder reactor in which the relatively abundant U-238 would be placed near a core of fissionable material. The reactor would be a fast reactor; that is, neutrons would not be moderated, as were most wartime reactors. Thus, the large number of neutrons emitted in fast neutron fission would hit the U-238 and create ''extra'' fissionable material

  17. Liquid Metal Fast Breeder Reactor Program (LMFBR): facility profiles

    International Nuclear Information System (INIS)

    A description is presented of the experimental test facilities involved in the conduct of the LMFBR research and development program. Existing facilities and those under construction or authorized as of October 1975 are described. Each profile presents brief descriptions of the overall facility and its test area and data relating to its experimental and testing capability. Introductory material for each section includes site and facility maps and an alphabetical list of the profiles contained in the section. A glossary of acronyms and letter designations in common usage to describe organizations, reactor and test facilities, components, etc. involved in the LMFBR program is included. Alphabetical, organizational, and programmatic indexes are provided as a convenient method of identifying the facilities with their locations and with their principal uses in the LMFBR program

  18. The long-term future for civilian nuclear power generation in France: The case for breeder reactors. Breeder reactors: The physical and physical chemistry parameters, associate material thermodynamics and mechanical engineering: Novelties and issues; La production d'electricite d'origine nucleaire en France, dans le futur a long terme: Le cas des surgenerateurs: Les reacteurs nucleaires surgenerateurs: Les parametres physique et physico-chimiques, la thermodynamique associee des materiaux et de l'ingenierie mecanique: Nouveautes et options

    Energy Technology Data Exchange (ETDEWEB)

    Dautray, R. [Academie des sciences, 23, quai de Conti, 75270 Paris cedex 06 (France)

    2011-06-15

    The author gives a summary overview of the knowledge base acquired since the first breeder reactors became operational in the fifties. Neutron transport theory, thermal phenomena, reactor core cooling, various coolants used and envisioned for this function, fuel fabrication from separated materials, main equipment (pumps, valves, heat exchanges...) have now attained maturity, sufficient to implement sodium cooling circuits. However, the use of metallic sodium still raises certain severe questions in terms of safe handling and security considerations. The structural components, both inside the reactor core and outside (i.e. heat exchangers) are undergoing in-depth research so as to last longer. The fuel cycle, notably the re-fabrication of fuel elements and fertile elements, the case of transuranic elements, etc., call for studies into radiation induced phenomena, chemistry separation, separate or otherwise treatments for materials that have different radioactive, physical, thermodynamical, chemical and biological properties. The concerns that surround the definitive disposal of certain radioactive wastes could be qualitatively improved with respect to the pressurized water reactors (PWRs) in service today. Lastly, the author notes that breeder reactors eliminate the need for an isotope separation facility, and this constitutes a significant contribution to contain nuclear proliferation. France was in the forefront of nuclear breeder power generation science, technological research and also in the knowledge base related to breeder reactors. It is in the country's interest to pursue these efforts. (author)

  19. An estimate of the radiological consequence of notional accidental releases of radioactivity from a fast breeder reactor

    International Nuclear Information System (INIS)

    In this report an assessment is made of the radiological consequences of notional accidental releases of activity from a fast breeder reactor under certain circumstances. It was prepared under contract to the Nuclear Installations Inspectorate (Health and Safety Executive) to assist them in making a preliminary safety assessment of fast reactors. The range of releases considered in the report was specified by the Inspectorate and comprises the vaporisation and release of varying fractions of the core of a 1300 MW(e) reactor. Two cases are evaluated depending on assumptions relating to the remainder of the core. No attempt is made to assign any probability to the occurrence of a given release; the report provides no more than a part of the information necessary for a safety assessment and is to be considered only within this limited context. The subject is dealt with under the following headings: introduction; parameters used in the assessment; atmospheric dispersion; pathways of exposure and dosimetric models; doses associated with the release of one tonne of fuel; biological effects; consequences of releases in which the nuclide composition differs from that in the fuel. (U.K.)

  20. ULOF transient behaviour of metal-fuelled fast breeder reactor cores as a function of core size and perturbation methods

    International Nuclear Information System (INIS)

    Highlights: • Metal fuel FBR safety can be assessed by its response to unprotected transients. • Safety during unprotected loss of flow accident (ULOF) is important for FBR cores. • ULOF analyses are carried out as a function of core size and perturbation method. • Smaller metal cores are found to be safer with respect to the ULOF accidents. • 1st order perturbation method gives conservative results in an ULOF accident. - Abstract: The safety behaviour of metal-fuelled fast breeder reactor cores may be assessed by their transient behaviour during anticipated unprotected transients. Out of such transients, unprotected loss of flow accident (ULOFA) has been recognized as an event important for determining reactor safety due to the positive sodium void coefficient of reactivity and the remote possibility of complete power failure as initiator. Reactor safety under ULOFA condition is particularly based on the inherent feedbacks, which is calculated using the removal worths and Doppler constants. As the removal worth is a strong function of reactor size, ULOF analyses are carried out in three different reactor size viz. 120 MWe, 500 MWe and 1000 MWe. The study reveals that smaller metal cores are safer than larger cores with respect to the ULOF accidents in the pre-disassembly phase. The present study also shows that the use of exact perturbation based reactivity worths introduce no significant changes in the safety behaviour of metal fuel reactor compared to that with the use of first order perturbation worths in pre-disassembly phase. The first order approximation is found to be valid as the expansion of materials in the core during ULOFA is small before the core enters the disassembly phase

  1. ULOF transient behaviour of metal-fuelled fast breeder reactor cores as a function of core size and perturbation methods

    Energy Technology Data Exchange (ETDEWEB)

    Riyas, A., E-mail: rias@igcar.gov.in [111B, CDO, Reactor Physics Division, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, Tamil Nadu (India); Mohanakrishnan, P. [Adjunct Professor, Manipal University, Manipal (India)

    2014-10-15

    Highlights: • Metal fuel FBR safety can be assessed by its response to unprotected transients. • Safety during unprotected loss of flow accident (ULOF) is important for FBR cores. • ULOF analyses are carried out as a function of core size and perturbation method. • Smaller metal cores are found to be safer with respect to the ULOF accidents. • 1st order perturbation method gives conservative results in an ULOF accident. - Abstract: The safety behaviour of metal-fuelled fast breeder reactor cores may be assessed by their transient behaviour during anticipated unprotected transients. Out of such transients, unprotected loss of flow accident (ULOFA) has been recognized as an event important for determining reactor safety due to the positive sodium void coefficient of reactivity and the remote possibility of complete power failure as initiator. Reactor safety under ULOFA condition is particularly based on the inherent feedbacks, which is calculated using the removal worths and Doppler constants. As the removal worth is a strong function of reactor size, ULOF analyses are carried out in three different reactor size viz. 120 MWe, 500 MWe and 1000 MWe. The study reveals that smaller metal cores are safer than larger cores with respect to the ULOF accidents in the pre-disassembly phase. The present study also shows that the use of exact perturbation based reactivity worths introduce no significant changes in the safety behaviour of metal fuel reactor compared to that with the use of first order perturbation worths in pre-disassembly phase. The first order approximation is found to be valid as the expansion of materials in the core during ULOFA is small before the core enters the disassembly phase.

  2. Experience in the maintenance of sodium systems of fast breeder test reactor

    International Nuclear Information System (INIS)

    The fast breeder test reactor (FBTR) is a loop type sodium cooled fast reactor located at Kalpakkam in India and that has been operating for 25 years. The reactor has been operated up to a power level of 18.6 MWt with a sodium outlet temperature of 482 C. degrees. Several modifications were carried out in the sodium systems to improve the plant performance. During the course of operation of the reactor, a number of sodium laden components like pumps, valves, cold traps, rupture disks, level probes, shielding plugs, control rod drive mechanisms, experimental assemblies, piping... were removed for various maintenance, modification and replacement jobs which has given the operators a valuable experience in handling large scale sodium systems. This paper details the special procedures followed during the handling of active and inactive sodium laden components

  3. Safety requirements expected to the prototype fast breeder reactor 'Monju'

    International Nuclear Information System (INIS)

    In July 2013, Nuclear Regulation Authority (NRA) has enforced new regulatory requirements in consideration of severe accidents for the commercial light water reactors (LWR) and also prototype power generation reactors such as the sodium-cooled fast reactors (SFR) of 'Monju' based on TEPCO Fukushima Daiichi nuclear power plant accident (hereinafter referred to as '1F accident') occurred in March 2011. Although the regulatory requirements for SFR will be revised by NRA with consideration for public comments, Japan Atomic Energy Agency (JAEA) set up 'Advisory Committee on Monju Safety Requirements' consisting of fast breeder reactor (FBR) and safety assessment experts in order to establish original safety requirements expected to the prototype FBR 'Monju' considering severe accidents with knowledge from JAEA as well as scientific and technical insights from the experts. This report summarizes the safety requirements expected to Monju discussed by the committee. (author)

  4. Lessons learned from the licensing process for the Clinch River Breeder Reactor Plant

    International Nuclear Information System (INIS)

    This paper presents the experience of licensing a specific liquid-metal fast breeder reactor (LMFBR), the Clinch River Breader Reactor Plant (CRBRP). It was a success story in that the licensing process was accomplished in a very short time span. The actions of the applicant and the actions of the US Nuclear Regulatory Commission (NRC) in response are presented and discussed to provide guidance to future efforts to license unconventional reactors. The history is told from the perspective of the authors. As such, some of the reasons given for success or lack of success are subjective interpretations. Nevertheless, the authors' positions provided them an excellent viewpoint to make these judgements. During the second phase of the licensing process, they were the CRBRP Technical Director and the Licensing Manager, respectively, for the Westinghouse Electric Corporation, the prime contractor for the reactor plant

  5. Status of national programmes on fast breeder reactors. Twenty-fifth annual meeting of the International Working Group on Fast Reactors. Summary report. Working material

    International Nuclear Information System (INIS)

    At present nuclear power accounts for approximately 17% of total electricity generation worldwide. Given continuing population growth and the needs of the third world and developing countries to improve their economic performance and standard of living, energy demand is expected to continue to grow through the 21st century. The proportion of energy supplied as electricity is also expected to continue to increase. Although fossil fuelled electricity generation is the option preferred by several countries for the short term, there are rising concerns over climatic consequences caused by extended burning of fossil fuels as a result of the demands of a fast expanding world population. In this situation nuclear electricity will become more and more important and the known reserves of uranium would be consumed quite quickly by thermal reactors. It would be possible to sustain a large nuclear programme only by introducing fast reactors. One can conclude that there are strategic reasons for pursuing the development of fast breeder reactors. It will become desirable essential, to have this technology available for introduction. The experience of the various prototypes presently in operation has confirmed the operability and benign characteristics of the LMFR and has given ground for confidence in the future. Current fast reactor designs offer very large margins of safety and by virtue of redundant and diverse safety systems the potential for an energetic core disruptive accident or for fast reactor core meltdown has been essentially eliminated. Several international forums reviewed the current trends in the fast reactor development. The view was reaffirmed that fast breeder reactors still remain the most practical tool for effective utilization of uranium resources for the future energy needs. Achievement of competitiveness with LMRs is still the first priority condition for the future deployment of this type of reactor. The recycling of plutonium into LMFBRs would allow

  6. Effects of nuclear island connected buildings on seismic behaviour of reactor internals in a pool type fast breeder reactor

    International Nuclear Information System (INIS)

    The seismic analysis of reactor assembly housing the primary circuit of a typical 500 MWe capacity pool type fast breeder reactor (PFBR) is reported. The reactor assembly is supported on the reactor vault within the nuclear island connected buildings (NICB). The seismic responses, viz. critical displacements, sloshing heights, stresses and strain energy values in the vessels are determined for the reactor assembly by detailed finite element analysis including the fluid-structure interaction and sloshing effects. Analysis is carried out to quantify the effects of inter-connection of the reactor vault with the adjacent buildings under the assumptions that the reactor vault along with reactor assembly is: (1) an isolated structural system from the adjacent buildings within reactor containment building (RCB) and (2) connected with the adjacent civil structures through floor slabs. Analysis indicates that, by inter-connecting the vault with the NICB, there are overall increases of all the governing parameters which decide the seismic design criteria. The significant effects are increases of: (1) radial and axial displacements of core top and absorber rods and vertical accelerations of core subassemblies which are of concern to reactor safety, (2) primary membrane stress intensities for the inner vessel and (3) strain energies developed at the critical portions which can enhance the buckling risks of main vessel, inner vessel and thermal baffles. Hence, it is preferable to isolate the reactor vault, directly constructing from the base raft without inter-connecting it with the NICB, from the seismic loading considerations

  7. Analysis for mechanical consequences of a core disruptive accident in Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    The mechanical consequences of a core disruptive accident (CDA) in a fast breeder reactor are described. The consequences are development of deformations and strains in the vessels, intermediate heat exchangers (IHX) and decay heat exchangers (DHX), impact of sodium slug on the bottom surface of the top shield, sodium release to reactor containment building through top shield penetrations, sodium fire and consequent temperature and pressure rise in reactor containment building (RCB). These are quantified for 500 MWe Prototype Fast Breeder Reactor (PFBR) for a CDA with 100 MJ work potential. The results are validated by conducting a series of experiments on 1/30 and 1/13 scaled down models with increasing complexities. Mechanical energy release due to nuclear excursion is simulated by chemical explosion of specially developed low density explosive charge. Based on these studies, structural integrity of primary containment, IHX and DHX is demonstrated. The sodium release to RCB is 350 kg which causes pressure rise of 12 kPa in RCB. (author)

  8. Tentative design-philosophy for bellows in sodium cooled fast breeder reactors pipings

    International Nuclear Information System (INIS)

    Expansion joints have proved to be reliable components, when properly designed and realized, in normal industrial equipment. But nevertheless bellows have not been employed widely in nuclear reactors and almost not in sodium cooled fast breeder reactors, where use of expansion-joints could considerably shorten the length of pipelines and, in consequence, lower the cost of the power plant. In the framework of its research and development program on fast reactors the French Atomic Energy.Commission, in cooperation with the industry, develops guidelines, backed up by experiments, to allow a safe design of pipe-lines and compensating-devices. The main points of these guidelines are discussed in this paper with the understanding, that they are tentative rules subject to changes. The guidelines are a complement to existing rules, like ASME - Code III, Code Case 1481, standards of the EJMA Preliminary Draft for Code Case Class I, Expansion Joints in Piping systems and suppliers' rules for the special case of application to sodium cooled fast breeder reactors. Relatively small diameters and easily accessible expansion joints, on control rods and valves for example, are not concerned. These guidelines do not apply to the bellows which are used as an integral part of a component

  9. Canadian power reactor fuel

    International Nuclear Information System (INIS)

    The following subjects are covered: the basic CANDU fuel design, the history of the bundle design, the significant differences between CANDU and LWR fuel, bundle manufacture, fissile and structural materials and coolants used in the CANDU fuel program, fuel and material behaviour, and performance under irradiation, fuel physics and management, booster rods and reactivity mechanisms, fuel procurement, organization and industry, and fuel costs. (author)

  10. Thermochemical study of material compatibility for sodium cooled fast breeder reactor application

    International Nuclear Information System (INIS)

    In fast breeder reactors, liquid sodium is preferred as a coolant due to its high thermal conductivity, high specific heat, low viscosity, wide liquid range, remarkable thermal stability. However, it must be in the pure form to be compatible with structural materials in which chemical compounds of Austenitic stainless used include carbon, chromium, nickel, molybdenum, iron, niobium, zirconium and so on. Traces of impurities play an important role in corrosion, mass transport loops of the reactor. Corrosion of structural materials in liquid sodium is deeply affected by the oxygen concentration. Some of these corrosion products which find their way into sodium can cause risk when they deposit on parts like heat exchangers and pumps, which have to be periodically maintained. Thus one must not only control and monitor the oxygen impurity level, but also understand the mechanism of the chemical reaction in the reactor. In this way, thermodynamic approach is obtained by analyzing compatibility of chemical compounds of structural materials with liquid sodium

  11. Implementation of multivariable control techniques with application to Experimental Breeder Reactor II

    International Nuclear Information System (INIS)

    After several successful applications to aerospace industry, the modern control theory methods have recently attracted many control engineers from other engineering disciplines. For advanced nuclear reactors, the modern control theory may provide major advantages in safety, availability, and economic aspects. This report is intended to illustrate the feasibility of applying the linear quadratic Gaussian (LQG) compensator in nuclear reactor applications. The LQG design is compared with the existing classical control schemes. Both approaches are tested using the Experimental Breeder Reactor 2 (EBR-2) as the system. The experiments are performed using a mathematical model of the EBR-2 plant. Despite the fact that the controller and plant models do not include all known physical constraints, the results are encouraging. This preliminary study provides an informative, introductory picture for future considerations of using modern control theory methods in nuclear industry. 10 refs., 25 figs

  12. The fast breeder reactor Rapsodie (1962); Le reacteur rapide surregenerateur rapsodie (1962)

    Energy Technology Data Exchange (ETDEWEB)

    Vautrey, L.; Zaleski, C.P. [Commissariat a l' Energie Atomique, Cadarache (France). Centre d' Etudes Nucleaires

    1962-07-01

    In this report, the authors describe the Rapsodie project, the French fast breeder reactor, as it stands at construction actual start-up. The paper provides informations about: the principal neutronic and thermal characteristics, the reactor and its cooling circuits, the main handling devices of radioactive or contaminated assemblies, the principles and means governing reactor operation, the purposes and locations of miscellaneous buildings. Rapsodie is expected to be critical by 1964. (authors) [French] Dans ce rapport, les auteurs font le point du projet RAPSODIE (reacteur francais surregenerateur a neutrons rapides), au moment du debut effectif de sa construction. On y trouvera decrits: les principales caracteristiques neutroniques et thermiques, le bloc pile et les circuits de refroidissement, les principaux moyens de manutention des ensembles actifs ou contamines, les principes et les moyens qui regissent la conduite du reacteur, les fonctions et l'implantation des divers batiments. La divergence de RAPSODIE est prevue pour 1964. (auteurs)

  13. Conjugate heat transfer analysis of multiple enclosures in prototype fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Velusamy, K.; Balaubramanian, V.; Vaidyanathan, G.; Chetal, S.C. [Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

    1995-09-01

    Prototype Fast Breeder Reactor (PFBR) is a 500 MWe sodium cooled reactor under design. The main vessel of the reactor serves as the primary boundary. It is surrounded by a safety vessel which in turn is surrounded by biological shield. The gaps between them are filled with nitrogen. Knowledge of temperature distribution prevailing under various operating conditions is essential for the assessment of structural integrity. Due to the presence of cover gas over sodium free level within the main vessel, there are sharp gradients in temperatures. Also cover gas height reduces during station blackout conditions due to sodium level rise in main vessel caused by temperature rise. This paper describes the model used to analyse the natural convection in nitrogen, conduction in structures and radiation interaction among them. Results obtained from parametric studies for PFBR are also presented.

  14. Review of ORNL-TSF shielding experiments for the gas-cooled Fast Breeder Reactor Program

    Energy Technology Data Exchange (ETDEWEB)

    Abbott, L.S.; Ingersoll, D.T.; Muckenthaler, F.J.; Slater, C.O.

    1982-01-01

    During the period between 1975 and 1980 a series of experiments was performed at the ORNL Tower Shielding Facility in support of the shield design for a 300-MW(e) Gas Cooled Fast Breeder Demonstration Plant. This report reviews the experiments and calculations, which included studies of: (1) neutron streaming in the helium coolant passageways in the GCFR core; (2) the effectiveness of the shield designed to protect the reactor grid plate from radiation damage; (3) the adequacy of the radial shield in protecting the PCRV (prestressed concrete reactor vessel) from radiation damage; (4) neutron streaming between abutting sections of the radial shield; and (5) the effectiveness of the exit shield in reducing the neutron fluxes in the upper plenum region of the reactor.

  15. Clinch River Breeder Reactor Plant steam generator: FEW tube test model post test examination

    International Nuclear Information System (INIS)

    The Steam Generator Few Tube Test (FTT) is part of an extensive testing program being carried out in support of the Clinch River Breeder Reactor Plant (CRBRP) steam generator design. The testing of full-length seven-tube evaporator and three-tube superheater models of the CRBRP design was conducted to provide steady-state thermal/hydraulic performance data to full power per tube and to verify the absence of multi-year endurance problems. The problems encountered with the mechanical features of the FTT model design which led to premature test termination and the results of the post-test examination are described

  16. A supercritical steam cooled fast breeder reactor with negative reactivity characteristics against voiding and flooding

    International Nuclear Information System (INIS)

    The steam cooled fast breeder reactor with negative reactivity characteristic against voiding and flooding is feasible under the supercritical pressure. The breeding ratio is 1.04. A flat core with the zirconium hydride layer is adopted for mitigating the void reactivity. The thermal efficiency of the indirect cycle system is improved 9 % relatively from the current PWR's. The core should be cooled in 10 seconds after the large break loss of coolant accident (LOCA). The coast down time should be larger than 30 seconds to overcome the loss of flow (LOF) by the trip of all blowers. (author)

  17. Compendium of computer codes for the safety analysis of fast breeder reactors

    International Nuclear Information System (INIS)

    The objective of the compendium is to provide the reader with a guide which briefly describes many of the computer codes used for liquid metal fast breeder reactor safety analyses, since it is for this system that most of the codes have been developed. The compendium is designed to address the following frequently asked questions from individuals in licensing and research and development activities: (1) What does the code do. (2) To what safety problems has it been applied. (3) What are the code's limitations. (4) What is being done to remove these limitations. (5) How does the code compare with experimental observations and other code predictions. (6) What reference documents are available

  18. Large scale breeder reactor plant prototype mechanical pump conceptual design study

    Energy Technology Data Exchange (ETDEWEB)

    1976-07-01

    This final report is a complete conceptual design study of a mechanical pump for a large scale breeder reactor plant. The pumps are located in the cold leg side of the loops. This makes the net positive suction head available - NPSHA - low, and is, in fact, a major influencing factor in the design. Where possible, experience gained from the Clinch River Project and the FFTF is used in this study. Experience gained in the design, manufacturer, and testing of pumps in general and sodium pumps in particular is reflected in this report. The report includes estimated cost and time schedule for design, manufacture, and testing. It also includes a recommendation for development needs.

  19. Comparison of diffusion and transport theory analysis with experimental results in fast breeder test reactor

    International Nuclear Information System (INIS)

    A systematic analysis has been performed by 3 dimensional diffusion and transport methods to calculate the measured control rod worths and subassembly wise power distribution in fast breeder test reactor. Geometry corrections (rectangular to hexagonal and diffusion to transport corrections are estimated for multiplication factors and control rod worths. Calculated control rod worths by diffusion and transport theory are nearly the same and 10% above measured values. Power distribution in the core periphery is over predicted (15%) by diffusion theory. But, this over prediction reduces to 8% by use of the SN method. (authors). 9 refs., 4 tabs., 3 fig

  20. Optimization of the Westinghouse/Stone and Webster prototype large breeder reactor

    International Nuclear Information System (INIS)

    The optimization of the Westinghouse/Stone and Webster Prototype Large Breeder Reactor (PLBR) is described. This reactor plant, designed for ERDA and EPRI, resulted from optimization and tradeoffs on plant size, number of loops, steam cycles, system temperature, pump location, refueling concept, reactor shut-down system logic, control system logic, steam generating system, residual heat removal system, core arrangement and reactor vessel. The result is a three loop LMFBR rated at 1000 MWe gross with sodium entering the reactor vessel at 6500F (3450C) and leaving at 9500F) (5100C). The reactor vessel has a flat closure head on top, containing three rotating plugs on which are mounted the upper internals structure and the in-vessel transfer machine. The core has three radial layers of core material separating four radial blanket regions. An inclined refueling chute penetrates the reactor vessel. Plant efficiency of 37% is achieved with the use of once-through steam generators operating in the modified Sulzer mode, producing 2200 psi, 8500F (15.2 x 106 pa/4550C) steam. The residual heat removal system (RHRS) consists of three independent heat removal paths in which the intermediate sodium is cooled in air blast heat exchangers (author)

  1. Prototype fast breeder reactor main and safety vessel surfaces in-service inspection mobile robot

    International Nuclear Information System (INIS)

    Periodic inspection of Prototype Fast Breeder Reactor (PFBR) main vessel and safety vessel is important to assess their structural integrity and to take remedial measures, if needed. PFBR is a pool type reactor and a safety vessel is provided in the design, which envelops the main reactor vessel. As the reactor inside is inaccessible, inspection can only be performed from outside the main vessel. Division of Remote Handling and Robotics, Bhabha Atomic Research Centre (BARC) in collaboration with Indira Gandhi Centre for Atomic Research (IGCAR) is working on the design of a proto-type mobile robot that would do the inspection of reactor vessel surfaces through the annular gap using friction grip when the reactor is in shutdown condition. This mobile robot will be inserted through the access holes at the top of the reactor vault leading to the annular space and moves around the vessel, carrying visual camera, lighting system and ultrasonic testing modules as accessories and also positions and orients them to do ISI of main and safety vessel surfaces. The details of the configuration of the VENTURE, method of achieving mobility around the vessel for coverage and adaptation to the variation in annular gap and other salient design features required to perform the ISI are briefed in this paper. (author)

  2. Implications of nuclear physics in the development of Fast Breeder Reactors

    International Nuclear Information System (INIS)

    The purpose of this paper is to point out the involved aspects of nuclear physics in the calculation and design of the fast reactors. After a brief description of the advantages of using the fast reactors in the national economy, the national programs concerning this activity are presented. The structure and operation conditions of the liquid metal fast breeder reactor (LMFBR) are also reviewed. Then, the methods aimed to calculate the core, the burn-up, the reactor dynamics, the analysis of accidents, the shielding, as well as, the materials required in the fast reactor calculation, are shortly given. Further on, it deals with the nuclear data types connected to the fast reactor calculations, with accuracy requirements for nuclear data, as well as, with the present stage of nuclear data for fissile, fertile and structural materials. The requirements for new differential data measurements, new integral data and benchmark experiments are presented. Data adjustement methods are also summarized. Some aspects of the structural material behaviour in intense gamma radiation and neutron fields existing into a fast reactor are also presented in the last part of this paper. The concluding remarks are mentioned at the end of the paper. (author)

  3. Proposed method of the modeling and simulation of corrosion product behavior in the primary cooling system of fast breeder reactors

    International Nuclear Information System (INIS)

    Radioactive corrosion products (CP) are main cause of personal radiation exposure during maintenance without fuel failure in FBR plants. In order to establish the techniques of radiation dose estimation for worker in radiation-controlled area, Program SYstem for Corrosion Hazard Evaluation code 'PSYCHE' has been developed. The PSYCHE is based on the Solution-Precipitation model. The CP transfer calculation using the Solution-Precipitation model needs a fitting factor for the calculation of the precipitation of CP. This fitting factor must be determined based on the measured values in reactors that have operating experience. For this reason, the inability to make accurate predictions for reactor without measured values is a major issue. In this study, in addition to existing Solution-Precipitation model in PSYCHE, a transfer-model of CP species in particle form was applied to calculations of CP behavior in the primary cooling system of fast breeder reactor MONJU. Based on the calculated results, we estimated the contribution of CP deposition in the particle-form. It was suggested that the improved model including transfer-model of CP species in particle-form could be used for evaluation of CP transfer and radiation-source distribution in place of conventional Solution-Precipitation model with fitting factor in the PSYCHE. Moreover, it was predicted that CP particles would tend to be deposited in region with high-flow rate of coolant. (author)

  4. Fueling of tandem mirror reactors

    International Nuclear Information System (INIS)

    This paper summarizes the fueling requirements for experimental and demonstration tandem mirror reactors (TMRs), reviews the status of conventional pellet injectors, and identifies some candidate accelerators that may be needed for fueling tandem mirror reactors. Characteristics and limitations of three types of accelerators are described; neutral beam injectors, electromagnetic rail guns, and laser beam drivers. Based on these characteristics and limitations, a computer module was developed for the Tandem Mirror Reactor Systems Code (TMRSC) to select the pellet injector/accelerator combination which most nearly satisfies the fueling requirements for a given machine design

  5. Solid breeder blanket concepts

    International Nuclear Information System (INIS)

    An investigation is made of a mechanical concept for the blanket with solid breeders in view of the possible adaptation to power reactor. A special arrangement of the multiplier and breeder materials is developed to permit a further neutronic optimisation

  6. Reactivity control system of a passively safe thorium breeder pebble bed reactor

    International Nuclear Information System (INIS)

    Highlights: • A worth of over 15,000 pcm ensures achieving long-term cold shutdown in thorium PBR. • Control rod worth in side reflector is insufficient due to low-power breeder zone. • 20 control rods, just outside the driver zone, can achieve long-term cold shutdown. • BF3 gas can be inserted for reactor shutdown, but only in case of emergency. • Perturbation theory accurately predicts absorber gas worth for many concentrations. - Abstract: This work investigates the neutronic design of the reactivity control system for a 100 MWth passively safe thorium breeder pebble bed reactor (PBR), a conceptual design introduced previously by the authors. The thorium PBR consists of a central driver zone of 100 cm radius, surrounded by a breeder zone with 300 cm outer radius. The fissile content of the breeder zone is low, leading to low fluxes in the radial reflector region. Therefore, a significant decrease of the control rod worth at this position is anticipated. The reactivity worth of control rods in the side reflector and at alternative in-core positions is calculated using different techniques, being 2D neutron diffusion, perturbation theory and more accurate 3D Monte Carlo models. Sensitivity coefficients from perturbation theory provide a first indication of effective control rod positions, while the 2D diffusion models provide an upper limit on the reactivity worth achievable at a certain radial position due to the homogeneous spreading of the absorber material over the azimuthal domain. Three dimensional forward calculations, e.g. in KENO, are needed for an accurate calculation of the total control rod worth. The two dimensional homogeneous calculations indicate that the reactivity worth in the radial reflector is by far insufficient to achieve cold reactor shutdown, which requires a control rod worth of over 15 000 pcm. Three dimensional heterogeneous KENO calculations show that placing 20 control rods just outside the driver channel, between 100 cm and

  7. Level-2 PSA for the prototype fast breeder reactor MONJU applied to the accident management review

    International Nuclear Information System (INIS)

    An accident management guideline (AMG) of the prototype fast breeder reactor MONJU has been presented to Nuclear and Industry Safety Agency (NISA) of METI by Japan Atomic Energy Agency (JAEA) with an evaluation result of an effectiveness of the AMG by employing Level-1 and Level-2 PSAs. Japan Nuclear Energy Safety Organization (JNES) evaluated the three events - PLOHS, LORL and ATWS events - and scrutinized the results of the Level-2 PSA carried out by JAEA from the view point of an accident management (AM) review. Regarding ATWS events, we have carried out a qualitative evaluation of the results of JAEA's evaluation and carried out a quantitative evaluation of the containment failure frequency (CFF) in relation to Protected-Loss-of-Heat-Sink (PLOHS) and Loss-of-Reactor-Level (LORL) events. Evaluation of the containment failure probability CFF has been conducted based on the results of the Level-1 PSA by employing the code system developed by JNES. We conducted a close examination of the procedure that JAEA followed to evaluate CFFs in PLOHS and LORL events. It was confirmed that JAEA's Level-2 PSA quantified the phenomenal event trees was expanded in the three processes - the plant response process, the core damage process and the containment vessel response process - based on various analytical and experimental evidence and otherwise followed much the same basic evaluation procedures employed by JNES. As for PLOHS and LORL, quantitative evaluation of CFF was conducted according to the following procedures: Development of an event flow diagram, Development of a phenomenal event tree, Quantification of the phenomenal event tree, Evaluation of containment failure frequencies, and Evaluation of the effectiveness of the AM measures. In the evaluation of the PLOHS and LORL events, the following analytical codes were used; Plant dynamic characteristic analytical code (NALAP-II), Nuclear characteristics analytical system (ARCADIAN-FBR/MVP), Nuclear dynamics analysis code

  8. Nuclear power engineering development on the basis of new conceptions of nuclear reactor and fuel cycle

    International Nuclear Information System (INIS)

    One analyzes the status of nuclear power industry (NPI) and lists the excuses explaining the modest progress of NPI in contrast to the predicted one. It is shown that progress of NPI equivalent to the expansion of power consumers may be ensured by construction of large breeder NPPs. One lists the requirements for reactor and for fuel cycle technologies. The design of the BREST fast UN-PuN fuel and lead-cooling reactor enables to meet the listed requirements

  9. Concept for a small, colocated fuel cycle facility for oxide breeder fuels

    International Nuclear Information System (INIS)

    As part of a United States Department of Energy (USDOE) program to examine innovative liquid-metal reactor (LMR) system designs over the past three years, the Oak Ridge National Laboratory (ORNL) and the Westinghouse Hanford Company (WHC) collaborated on studies of mixed oxide fuel cycle options. A principal effort was an advanced concept for a small integrated fuel cycle colocated with a 1300-MW(e) reactor station. The study provided a scoping design, capital and operating cost estimates, and a basis on which to proceed with implementation of such a facility if future plans so dictate. The facility integrated reprocessing, waste management, and refabrication functions in a single facility of nominal 35-t/year capacity utilizing the latest technology developed in fabrication programs at WHC and in reprocessing at ORNL. The concept was based on many years of work at both sites and extensive design studies of prior years

  10. Review of uncertainty estimates associated with models for assessing the impact of breeder reactor radioactivity releases

    International Nuclear Information System (INIS)

    The purpose is to summarize estimates based on currently available data of the uncertainty associated with radiological assessment models. The models being examined herein are those recommended previously for use in breeder reactor assessments. Uncertainty estimates are presented for models of atmospheric and hydrologic transport, terrestrial and aquatic food-chain bioaccumulation, and internal and external dosimetry. Both long-term and short-term release conditions are discussed. The uncertainty estimates presented in this report indicate that, for many sites, generic models and representative parameter values may be used to calculate doses from annual average radionuclide releases when these calculated doses are on the order of one-tenth or less of a relevant dose limit. For short-term, accidental releases, especially those from breeder reactors located in sites dominated by complex terrain and/or coastal meteorology, the uncertainty in the dose calculations may be much larger than an order of magnitude. As a result, it may be necessary to incorporate site-specific information into the dose calculation under these circumstances to reduce this uncertainty. However, even using site-specific information, natural variability and the uncertainties in the dose conversion factor will likely result in an overall uncertainty of greater than an order of magnitude for predictions of dose or concentration in environmental media following shortterm releases

  11. Status of Fast Breeder Reactor Development in the Federal Republic of Germany, Belgium and the Netherlands

    International Nuclear Information System (INIS)

    In 1967 and 1968, the Federal Republic of Germany, the Kingdom of Belgium and the Kingdom of the Netherlands (''DeBeNe'') agreed to develop breeder reactors in a joint program. The following research organizations have taken part in this effort: - Kernforschungszentrum Karlsruhe (KfK) - Interatom, Bergisch Gladbach Alkem, Hanau - SCK/CEN, Mol - Belgonucleaire, Brussels - ECN, Petten - TNO, Apeldoorn - Neratoom, The Hague. The first three institutions mentioned above have been associated in the Entwicklungsgemeinschaft Schneller Brüter since 1977. KfK, INTERATOM, and the French Commissariat à l'Energie Atomique entered into contracts in 1977 about close cooperation in the fast breeder field, to which the Belgian and Dutch partners acceded. The results of activities carried out by the DeBeNe partners in 1987 have been compiled in this report. The report begins with a survey of the fast reactor plants, which is followed by an R&D summary. In an additional chapter, a survey is given of international cooperation in 1987

  12. Numerical simulation of convection of argon gas in fast breeder reactor

    International Nuclear Information System (INIS)

    Highlights: • Turbulent convective motion of Argon gas in the annulus of fast breeder reactor. • Circumferential temperature difference (CTD) is proportional to the Rayleigh Number. • A pair of ascending and descending rolls that move azimuthally in the annulus. • Observed flow reversals. • Temperature fluctuations decrease with the height of the annulus. - Abstract: In this paper, we present the results of numerical simulations of the turbulent convection in the Argon gas present in the annulus of a fast breeder reactor. We employ RANS scheme with k–∊ model and solve the equations using an open-source software OpenFOAM. The Rayleigh numbers Ra of our simulations lie in the range of 108 to 1010. We observe a pair of rolls with a hot plume rising from one end, and a cold plume descending from the opposite end of the annulus. This feature results because the aspect ratio of the geometry is near unity. We also find that the circumferential temperature difference (CTD) is proportional to Ra

  13. Review of uncertainty estimates associated with models for assessing the impact of breeder reactor radioactivity releases

    Energy Technology Data Exchange (ETDEWEB)

    Miller, C.; Little, C.A.

    1982-08-01

    The purpose is to summarize estimates based on currently available data of the uncertainty associated with radiological assessment models. The models being examined herein are those recommended previously for use in breeder reactor assessments. Uncertainty estimates are presented for models of atmospheric and hydrologic transport, terrestrial and aquatic food-chain bioaccumulation, and internal and external dosimetry. Both long-term and short-term release conditions are discussed. The uncertainty estimates presented in this report indicate that, for many sites, generic models and representative parameter values may be used to calculate doses from annual average radionuclide releases when these calculated doses are on the order of one-tenth or less of a relevant dose limit. For short-term, accidental releases, especially those from breeder reactors located in sites dominated by complex terrain and/or coastal meteorology, the uncertainty in the dose calculations may be much larger than an order of magnitude. As a result, it may be necessary to incorporate site-specific information into the dose calculation under these circumstances to reduce this uncertainty. However, even using site-specific information, natural variability and the uncertainties in the dose conversion factor will likely result in an overall uncertainty of greater than an order of magnitude for predictions of dose or concentration in environmental media following shortterm releases.

  14. Comparison of material property specifications of austenitic steels in fast breeder reactor technology

    International Nuclear Information System (INIS)

    Austenitic stainless steels are very widely used in components for European Fast Breeder Reactors. The Activity Group Nr.3 ''Materials'', within Working Group ''Codes and Standards'' of the Fast Reactor Co-Ordination Committee of the European Communities, has decided to initiate a study to compare the material property specifications of the austenitic stainless steel used in the European Fast Breeder Technology. Hence, this study would allow one to view rapidly the designation of a particular steel grade in different European countries and to compare given property values for a same grade. There were dissimilarities, differences or voids appear, it could lead to an attempt to complete and/or to uniformize the nationally given values, so that on a practical level interchangeability, availability and use ease design and construction work. A selection of the materials and of their properties has been made by the Working Group. Materials examined are Stainless Steel AISI 304, 304 L, 304 LN, 316, 316 L, 316 LN, 316''Ti stab.'', 316''Nb stab''., 321, 347

  15. Shippingport operations with the Light Water Breeder Reactor core. (LWBR Development Program)

    Energy Technology Data Exchange (ETDEWEB)

    Budd, W.A. (ed.)

    1986-03-01

    This report describes the operation of the Shippingport Atomic Power Station during the LWBR (Light Water Breeder Reactor) Core lifetime. It also summarizes the plant-oriented operations during the period preceding LWBR startup, which include the defueling of The Pressurized Water Reactor Core 2 (PWR-2) and the installation of the LWBR Core, and the operations associated with the defueling of LWBR. The intent of this report is to examine LWBR experience in retrospect and present pertinent and significant aspects of LWBR operations that relate primarily to the nuclear portion of the Station. The nonnuclear portion of the Station is discussed only as it relates to overall plant operation or to unusual problems which result from the use of conventional equipment in radioactive environments. 30 refs., 69 figs., 27 tabs.

  16. Liquid metal seal (LMS) - challenges for fast breeder test reactor (FBTR)

    International Nuclear Information System (INIS)

    In Fast Breeder Test reactor (FBTR), Liquid Metal Seal (LMS) is being used to maintain leak tightness between reactor vessel and rotating plugs. It is a eutectic mixture of 42% tin and 58% bismuth. This paper describes measurements of melting point of LMS using Differential Scanning Calorimeter (DSC), Make: Setaram; Model- 131 evo. The instrument was calibrated using Indium as standard with different heating rates, 5 °C/min, 10 °C/min, 15°C/min and 20 °C/min. The observed value of melting point was found to be in agreement with the literature value. The melting point of as received and used LMS (LMSH8, LMSH10 and LMSH12) from three locations of FBTR were studied using DSC with different heating rates as above. The results are presented and it can be clearly seen that LMS has undergone some modifications during the continuous usage in FBTR

  17. Low-order dynamic modeling of the Experimental Breeder Reactor II

    International Nuclear Information System (INIS)

    This report describes the development of a low-order, linear model of the Experimental Breeder Reactor II (EBR-II), including the primary system, intermediate heat exchanger, and steam generator subsystems. The linear model is developed to represent full-power steady state dynamics for low-level perturbations. Transient simulations are performed using model building and simulation capabilities of the computer software Matrixx. The inherently safe characteristics of the EBR-II are verified through the simulation studies. The results presented in this report also indicate an agreement between the linear model and the actual dynamics of the plant for several transients. Such models play a major role in the learning and in the improvement of nuclear reactor dynamics for control and signal validation studies. This research and development is sponsored by the Advanced Controls Program in the Instrumentation and Controls Division of the Oak Ridge National Laboratory. 17 refs., 67 figs., 15 tabs

  18. Single assembly preliminary analysis for horizontal seismic analysis on fast breeder reactor core

    International Nuclear Information System (INIS)

    Seismic analysis is one of important parts of fast breeder reactor (FBR) core design. It is necessary for structural integrity assessment and analysis of variation of reactivity under the earthquake. Moreover some important data for qualification of the scram capability of the control rods during the earthquake. Moreover some important data for qualification of the scram capability of the control rods during the earthquake could be provided. In the paper, FINAS, one finite element code developed by Japanese engineers, was used. The calculation model and method were studied on single assembly in China Experimental Fast Reactor (CEFR), as an example. Some preliminary analyses were carried out, which prepare for the seismic analysis on multiple assemblies in FBR core. (authors)

  19. Experimental Breeder Reactor II (EBR-II), instrumentation for core surveillance

    International Nuclear Information System (INIS)

    The paper describes the Experimental Breeder Reactor-2 (EBR-2), thermal-hydraulic testing on the facility, and features of EBR-2 subassembly design. It is reported that during 25 years of EBR-2 operation, several of original, non-replaceable flow-sensors and thermocouples have failed in the primary system, and that this has led to the development of new sensors. The conclusion is made that from test series of measurements of temperature and flow in subassemblies, EBR-2 calculations showed that the core could withstand a loss-of-flow without scram accident and a loss-of-heat sink without scram accident from full reactor power without core damage. 11 refs, 9 figs

  20. Theory, design, and operation of liquid metal fast breeder reactors, including operational health physics

    Energy Technology Data Exchange (ETDEWEB)

    Adams, S.R.

    1985-10-01

    A comprehensive evaluation was conducted of the radiation protection practices and programs at prototype LMFBRs with long operational experience. Installations evaluated were the Fast Flux Test Facility (FFTF), Richland, Washington; Experimental Breeder Reactor II (EBR-II), Idaho Falls, Idaho; Prototype Fast Reactor (PFR) Dounreay, Scotland; Phenix, Marcoule, France; and Kompakte Natriumgekuhlte Kernreak Toranlange (KNK II), Karlsruhe, Federal Republic of Germany. The evaluation included external and internal exposure control, respiratory protection procedures, radiation surveillance practices, radioactive waste management, and engineering controls for confining radiation contamination. The theory, design, and operating experience at LMFBRs is described. Aspects of LMFBR health physics different from the LWR experience in the United States are identified. Suggestions are made for modifications to the NRC Standard Review Plan based on the differences.

  1. Steam condenser optimization using Real-parameter Genetic Algorithm for Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Highlights: → We model design optimization of a vital reactor component using Genetic Algorithm. → Real-parameter Genetic Algorithm is used for steam condenser optimization study. → Comparison analysis done with various Genetic Algorithm related mechanisms. → The results obtained are validated with the reference study results. - Abstract: This work explores the use of Real-parameter Genetic Algorithm and analyses its performance in the steam condenser (or Circulating Water System) optimization study of a 500 MW fast breeder nuclear reactor. Choice of optimum design parameters for condenser for a power plant from among a large number of technically viable combination is a complex task. This is primarily due to the conflicting nature of the economic implications of the different system parameters for maximizing the capitalized profit. In order to find the optimum design parameters a Real-parameter Genetic Algorithm model is developed and applied. The results obtained are validated with the reference study results.

  2. Shippingport operations with the Light Water Breeder Reactor core. (LWBR Development Program)

    International Nuclear Information System (INIS)

    This report describes the operation of the Shippingport Atomic Power Station during the LWBR (Light Water Breeder Reactor) Core lifetime. It also summarizes the plant-oriented operations during the period preceding LWBR startup, which include the defueling of The Pressurized Water Reactor Core 2 (PWR-2) and the installation of the LWBR Core, and the operations associated with the defueling of LWBR. The intent of this report is to examine LWBR experience in retrospect and present pertinent and significant aspects of LWBR operations that relate primarily to the nuclear portion of the Station. The nonnuclear portion of the Station is discussed only as it relates to overall plant operation or to unusual problems which result from the use of conventional equipment in radioactive environments. 30 refs., 69 figs., 27 tabs

  3. Theory, design, and operation of liquid metal fast breeder reactors, including operational health physics

    International Nuclear Information System (INIS)

    A comprehensive evaluation was conducted of the radiation protection practices and programs at prototype LMFBRs with long operational experience. Installations evaluated were the Fast Flux Test Facility (FFTF), Richland, Washington; Experimental Breeder Reactor II (EBR-II), Idaho Falls, Idaho; Prototype Fast Reactor (PFR) Dounreay, Scotland; Phenix, Marcoule, France; and Kompakte Natriumgekuhlte Kernreak Toranlange (KNK II), Karlsruhe, Federal Republic of Germany. The evaluation included external and internal exposure control, respiratory protection procedures, radiation surveillance practices, radioactive waste management, and engineering controls for confining radiation contamination. The theory, design, and operating experience at LMFBRs is described. Aspects of LMFBR health physics different from the LWR experience in the United States are identified. Suggestions are made for modifications to the NRC Standard Review Plan based on the differences

  4. Experience of secondary cooling system modification at prototype fast breeder reactor MONJU (Translated document)

    International Nuclear Information System (INIS)

    The prototype fast breeder reactor MONJU has been shut down since the secondary sodium leak accident that occurred in December 1995. After the accident, an investigation into the cause and a comprehensive safety review of the plant were conducted, and various countermeasures for sodium leak were examined. Modification work commenced in September 2005. Since sodium, a chemically active material, is used as coolant in MONJU, the modification work required work methods suitable for the handling of sodium. From this perspective, the use of a plastic bag when opening the sodium boundary, oxygen concentration control in a plastic bag, slightly-positive pressure control of cover gas in the systems, pressing and cutting with a roller cutter to prevent the incorporation of metal fillings, etc. were adopted, with careful consideration given to experience and findings from previous modification work at the experimental fast reactor JOYO and plants abroad. Owing to these work methods, the modification work proceeded close to schedule without incident. (author)

  5. Equipment cell liners for liquid-metal-cooled fast breeder reactors

    International Nuclear Information System (INIS)

    Concepts and practices used in the design of equipment cell liners for liquid-metal-cooled fast breeder reactor (LMFBR) systems were surveyed to ascertain the manner by which the functional requirements were satisfied, the severity of sodium spills the liners were designed to accommodate, and the problems encountered in design and construction. The survey was limited to ''loop-type'' LMFBRs, with primary interest on recently constructed plants. Steel-lined concrete structures are discussed; cell-liner designs used in several LMFBR plants are described with particular emphasis on the Southwest Experimental Fast Oxide Reactor (SEFOR), which uses a fixed liner, and the Fast Flux Test Facility (FFTF), which uses a free-floating liner; and research and development work believed necessary to permit a rational and thorough assessment of cell-liner design concept is identified

  6. Significance of the SNR 300 fast breeder reactor in terms of research policy

    International Nuclear Information System (INIS)

    The publication consists of the following documents: (1) Compact version of the expert opinion on the benefit to be gained for the research policy of the FRG from the FBR prototype reactor station SNR 300. (2) Speech of the Federal Minister for Research and Technology, Dr. Heinz Riesenhuber, in the German Bundestag (September 22, 1988). (3) Survey of fast breeder reactor development and of the SNR 300. (4) Statement concerning the proposal to use the Kalkar nuclear power station (SNR 300) as a 'waste management facility' for plutonium and other actinides. (5) Reply of the Federal Government to an interpellation filed by the deputies Mr Wetzel, Mr Stratmann, Mrs Tauber, Dr. Daniels (Regensburg), and the parliamentary party of the Greens. (orig./UA)

  7. Design optimization of backup seal for sodium cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Highlights: ► Design arrived from fourteen geometric options by finite element analysis. ► Seal geometry, size, compression, contact pressure, stress and compression load optimized. ► Effects of reduced fluoroelastomer strength at 110 °C, strain rate and stress-softening incorporated. ► Ageing, friction, tolerances, batch-to-batch/production variations in fluoroelastomer considered. ► Procedure applicable to other elastomeric seals of Fast Breeder Reactors. -- Abstract: Design optimization of static, fluoroelastomer backup seals for the 500 MWe, Prototype Fast Breeder Reactor (PFBR) is depicted. 14 geometric variations of a solid trapezoidal cross-section were studied by finite element analysis (FEA) to arrive at a design with hollowness and double o-ring contours on the sealing face. The seal design with squeeze of 5 mm assures failsafe operation for at least 10 years under a differential pressure of 25 kPa and ageing influences of fluid (air), temperature (110 °C) and γ radiation (23 mGy/h) in reactor. Hybrid elements of 1 mm length, regular integration, Mooney–Rivlin material model and Poisson’s ratio of 0.493 were used in axisymmetric analysis scheme. Possible effects of reduced fluoroelastomer strength at 110 °C, ageing, friction, tolerances in reactor scale, testing conditions during FEA data generation and batch-to-batch/production variations in seal material were considered to ensure adequate safety margin at the end of design life. The safety margin and numerical prediction accuracy could be improved further by using properties of specimens extracted from seal. The approach is applicable to other low pressure, moderate temperature elastomeric sealing applications of PFBR, mostly operating under maximum strain of 50%.

  8. Hydrodynamic and elastoplastic structural analysis of fast breeder reactor core accident

    International Nuclear Information System (INIS)

    This paper describes the principles and examples of applications of an explicit Lagrangian coupled finite difference-finite element code HEMP-ESI developed in order to calculate the structural consequences of hypothetical core disruptive accidents (HCDA) in nuclear reactors. The explicit solution algorithm of the finite difference scheme used to discretize the hydrodynamic fluid domains is shown to be very similar to that used for the solution of the finite element discretized shell structures, hence permitting an easy and efficient coupling. Two examples of simulation show the applicability of the method to nuclear reactor core safety analysis (test problem). Core explosion in a loop-type reactor including a shell containment: the calculation shows the energy absorbing function of the shell and enables the evaluation of the forces acting on the reactor containment. Hypothetical Core Disruptive Accident in a fast breeder reactor: the calculation shows the main features of this accident: lifting of the liquid sodium above the explosion and impact on the cover head inducing upward deformations; radial outflow of the sodium which induces large deformations of the inner and outer shell; zones of compressive circumferential stresses in the main shell at the junction of the spherical head and the cylindrical part

  9. Development of remotely controlled in-service inspection equipment for fast breeder reactor vessels

    International Nuclear Information System (INIS)

    For the operation of fast breeder reactors, one of the most important aspects is the need to control the functioning of the components. It is a characteristic of FBRs that the reactor vessel and in particular the guard vessel operate under very severe conditions. Therefore, an improved remotely controlled inspection system would be needed. On the basis of its experience with light water reactors, Mitsubishi Heavy Industries (MHI) has developed versatile systems for in-service inspection (ISI) of the reactor vessel and its ancillary components. The paper describes what could become the most important part of the ISI system, namely a special mobile vehicle or robot, called the MOLE (Mitsubishi Original LocomobilE). This vehicle can run freely over the annulus sections of the reactor vessel and its guard vessel and can carry out various tests. The results so far have been satisfactory and have encouraged MHI to make further tests in order to confirm that the access to the vessel is sufficient for performing the necessary inspections. (author)

  10. FUSED REACTOR FUELS

    Science.gov (United States)

    Mayer, S.W.

    1962-11-13

    This invention relates to a nuciear reactor fuel composition comprising (1) from about 0.01 to about 50 wt.% based on the total weight of said composition of at least one element selected from the class consisting of uranium, thorium, and plutonium, wherein said eiement is present in the form of at least one component selected from the class consisting of oxides, halides, and salts of oxygenated anions, with components comprising (2) at least one member selected from the class consisting of (a) sulfur, wherein the sulfur is in the form of at least one entity selected irom the class consisting of oxides of sulfur, metal sulfates, metal sulfites, metal halosulfonates, and acids of sulfur, (b) halogen, wherein said halogen is in the form of at least one compound selected from the class of metal halides, metal halosulfonates, and metal halophosphates, (c) phosphorus, wherein said phosphorus is in the form of at least one constituent selected from the class consisting of oxides of phosphorus, metal phosphates, metal phosphites, and metal halophosphates, (d) at least one oxide of a member selected from the class consisting of a metal and a metalloid wherein said oxide is free from an oxide of said element in (1); wherein the amount of at least one member selected from the class consisting of halogen and sulfur is at least about one at.% based on the amount of the sum of said sulfur, halogen, and phosphorus atom in said composition; and wherein the amount of said 2(a), 2(b) and 2(c) components in said composition which are free from said elements of uranium, thorium, arid plutonium, is at least about 60 wt.% based on the combined weight of the components of said composition which are free from said elements of uranium, thorium, and plutonium. (AEC)

  11. Numerical analysis of grid plate melting after a severe accident in a Fast-Breeder Reactor (FBR)

    Indian Academy of Sciences (India)

    A Jasmin Sudha; K Velusamy

    2013-12-01

    Fast breeder reactors (FBRs) are provided with redundant and diverse plant protection systems with a very low failure probability (<10-6/reactor year), making core disruptive accident (CDA), a beyond design basis event (BDBE). Nevertheless, safety analysis is carried out even for such events with a view to mitigate their consequences by providing engineered safeguards like the in-vessel core catcher. During a CDA, a significant fraction of the hot molten fuel moves downwards and gets relocated to the lower plate of grid plate. The ability of this plate to resist or delay relocation of core melt further has been investigated by developing appropriate mathematical models and translating them into a computer code HEATRAN-1. The core melt is a time dependent volumetric heat source because of the radioactive decay of the fission products which it contains. The code solves the nonlinear heat conduction equation including phase change. The analysis reveals that if the bottom of grid plate is considered to be adiabatic, melt-through of grid plate (i.e., melting of the entire thickness of the plate) occurs between 800 s and 1000 s depending upon the initial conditions. Knowledge of this time estimate is essential for defining the initial thermal load on the core catcher plate. If heat transfer from the bottom of grid plate to the underlying sodium is taken into account, then melt-through does not take place, but the temperature of grid plate is high enough to cause creep failure.

  12. Fuels for Canadian research reactors

    International Nuclear Information System (INIS)

    This paper includes some statements and remarks concerning the uranium silicide fuels for which there is significant fabrication in AECL, irradiation and defect performance experience; description of two Canadian high flux research reactors which use high enrichment uranium (HEU) and the fuels currently used in these reactors; limited fabrication work done on Al-U alloys to uranium contents as high as 40 wt%. The latter concerns work aimed at AECL fast neutron program. This experience in general terms is applied to the NRX and NRU designs of fuel

  13. Reactor core and fueling method

    International Nuclear Information System (INIS)

    When MOX fuel assemblies are used in a portion of fuel assembly of a BWR type nuclear reactor, neutron spectra are hardened due to the difference of the nuclear property between uranium and plutonium. As a result, the reactivity controllability of burnable poisons such as gadolinia is lowered, and the multiplication factor of the MOX fuel assembly at the initial stage of burning is increased greater than that of an uranium fuel assembly, to reduce thermal margin and reactor shutdown margin. Then, in the present invention, fresh fuel assemblies containing plutonium are disposed in a first region at the second layer from the outermost circumference of the reactor core and in a second region in adjacent with a control cell. Since the MOX fuel assemblies with increasing reactivity are disposed in the first and the second regions of small neutron importance, the power at the periphery of the reactor core and the circumference of the control cell can be kept substantially constant throughout the operation period. Further, satisfactory reactor operation can be kept without causing excess distortion of power distribution. (N.H.)

  14. Gaseous fuel nuclear reactor research

    Science.gov (United States)

    Schwenk, F. C.; Thom, K.

    1975-01-01

    Gaseous-fuel nuclear reactors are described; their distinguishing feature is the use of fissile fuels in a gaseous or plasma state, thereby breaking the barrier of temperature imposed by solid-fuel elements. This property creates a reactor heat source that may be able to heat the propellant of a rocket engine to 10,000 or 20,000 K. At this temperature level, gas-core reactors would provide the breakthrough in propulsion needed to open the entire solar system to manned and unmanned spacecraft. The possibility of fuel recycling makes possible efficiencies of up to 65% and nuclear safety at reduced cost, as well as high-thrust propulsion capabilities with specific impulse up to 5000 sec.

  15. Reactor fuel assemblies

    International Nuclear Information System (INIS)

    A description is given of an improved spacer grid for a nuclear fuel assembly comprising fuel rods in a matrix wherein each rod is adapted to be enclosed by a spacer ''cell'' for positioning thereof relative to adjacent rods in the fuel assembly. 7 claims, 12 drawing figures

  16. Core design of heavy water cooled thorium breeder reactor with negative void reactivity and improved breeding performance

    International Nuclear Information System (INIS)

    A core of heavy water cooled thorium breeder reactor that produces 3.5 GWt energy using Th-233U oxide fuel has been studied to depict a concrete design specification. In order to improve the breeding performance compared to that of our previous study, one of key parameters in core design: moderator to fuel volume ratio (MFR) is re-surveyed. By reducing MFR from 1.0 to 0.6, the swing of keff during a cycle is considerably flattened, keeping negative void coefficient. The batch number is 3 and the refueling scheme employs out-in method to limit the radial power peaking factor less than 1.3. Due to efficient internal conversion, the reactivity of the core slightly increases with burnup, so that the cycle length is extended up to 1,300 days. Consequently, high averaged burnup of 80 GWd/t and breeding ratio of 1.07 at middle of cycle is achieved without any blankets. The number of control rods made of B4C is 19 and the total reactivity worth is -6.5% dk/k. The present core uses Zircaloy-4 as cladding material, the fast neutron fluence at EOC (End Of Cycle), however, exceeds its limit due to hard spectrum and long cycle length. As a part of future study, design will be further explored considering cladding integrity. (authors)

  17. Micro-structural study and Rietveld analysis of fast reactor fuels: U-Mo fuels

    Science.gov (United States)

    Chakraborty, S.; Choudhuri, G.; Banerjee, J.; Agarwal, Renu; Khan, K. B.; Kumar, Arun

    2015-12-01

    U-Mo alloys are the candidate fuels for both research reactors and fast breeder reactors. In-reactor performance of the fuel depends on the microstructural stability and thermal properties of the fuel. To improve the fuel performance, alloying elements viz. Zr, Mo, Nb, Ti and fissium are added in the fuel. The first reactor fuels are normally prepared by injection casting. The objective of this work is to compare microstructure, phase-fields and hardness of as-cast four different U-Mo alloy (2, 5, 10 and 33 at.% Mo) fuels with the equilibrium microstructure of the alloys. Scanning electron microscope with energy dispersive spectrometer and optical microscope have been used to characterize the morphology of the as-cast and annealed alloys. The monoclinic α'' phase in as-cast U-10 at.% Mo alloy has been characterized through Rietveld analysis. A comparison of metallographic and Rietveld analysis of as-cast (dendritic microstructure) and annealed U-33 at.% Mo alloy, corresponding to intermetallic compound, has been reported here for the first time. This study will provide in depth understanding of microstructural and phase evolution of U-Mo alloys as fast reactor fuel.

  18. Accelerator breeder with uranium, thorium target

    International Nuclear Information System (INIS)

    An accelerator breeder, that uses a low-enriched fuel as the target material, can produce substantial amounts of fissile material and electric power. A study of H2O- and D2O-cooled, UO2, U, (depleted U), or thorium indicates that U-metal fuel produces a good fissile production rate and electrical power of about 60% higher than UO2 fuel. Thorium fuel has the same order of magnitude as UO2 fuel for fissile-fuel production, but the generating electric power is substantially lower than in a UO2 reactor. Enriched UO2 fuel increases the generating electric power but not the fissile-material production rate. The Na-cooled breeder target has many advantages over the H2O-cooled breeder target

  19. Experimental research of local hydrodynamic characteristics of fast reactor fuel assembly peripheral zone. 4

    International Nuclear Information System (INIS)

    Measurements were made of shear stress distribution and the velocity field of an aerodynamic model of the fast breeder reactor fuel assembly periphery. The effect was studied of a 50% disturbance of the geometry of a corner rod in a fuel assembly as against normal geometry. The coefficient of friction in the channel was assessed in dependence on the Reynolds number. The distribution of shear stresses in the walls of the fuel assembly and on spacers is graphically represented. (M.D.)

  20. Development of advanced in-service inspection technologies for Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Periodic in-service inspection (ISI) is a mandatory requirement for all nuclear power facilities due to the steadily increasing obligations to provide evidence that plant and equipment is consistently satisfying safety and integrity requirements. Periodic ISI will ensure that service-induced defects and abnormalities are detected and remedied at an early stage. Further, the ISI data will be immensely useful for arriving at appropriate decision on the life extension of the plants. In view of the high radiation and temperature prevailing in the nuclear plants, customized remote handling technologies coupled with robotic devices are a must for carrying out ISI of critical components in these plants. The 500 MWe Prototype Fast Breeder Reactor(PFBR), which is under construction at Kalpakkam, Tamil Nadu, India is a pool type reactor comprising of main vessel, safety vessel, and reactor roof structure confining the primary coolant and cover gas with any associated radioactivity. PFBR is designed with provisions for continuous monitoring and surveillance of the structures forming the primary containment. Supplementing the continuous surveillance and monitoring, the integrity of the reactor vessels has to be assessed periodically by ISI during the reactor shutdown conditions and a comprehensive ISI system capable of functioning at 150 deg C has been conceived and formulated to meet this mandatory requirement. The comprehensive ISI system comprises of two remote-controlled four-wheeled robotic devices with non-destructive examination modules for volumetric and visual examinations and each device has a location-specific inspection requirements. The remote-controlled devices can move around the annulus of the main vessel and safety vessel with inspection modules for carrying out the inspection. An essential feature of the ISI of the PFBR vessels is the formulation and establishment of permanent reference markers on the MV and SV to identify the location of the ISI device in

  1. Tritium trapping in silicon carbide in contact with solid breeder under high flux isotope reactor irradiation

    International Nuclear Information System (INIS)

    The trapping of tritium in silicon carbide (SiC) injected from ceramic breeding materials was examined via tritium measurements using imaging plate (IP) techniques. Monolithic SiC in contact with ternary lithium oxide (lithium titanate and lithium aluminate) as a ceramic breeder was irradiated in the High Flux Isotope Reactor (HFIR) in Oak Ridge, Tennessee, USA. The distribution of photo-stimulated luminescence (PSL) of tritium in SiC was successfully obtained, which separated the contribution of 14C β-rays to the PSL. The tritium incident from ceramic breeders was retained in the vicinity of the SiC surface even after irradiation at 1073 K over the duration of ∼3000 h, while trapping of tritium was not observed in the bulk region. The PSL intensity near the SiC surface in contact with lithium titanate was higher than that obtained with lithium aluminate. The amount of the incident tritium and/or the formation of a Li2SiO3 phase on SiC due to the reaction with lithium aluminate under irradiation likely were responsible for this observation

  2. Development of advanced blanket materials for solid breeder blanket of fusion reactor

    International Nuclear Information System (INIS)

    The design of advanced solid breeding blanket in the DEMO reactor requires the tritium breeder and neutron multiplier that can withstand the high temperature and high fluence, and the development of such as advanced blanket materials has been carried out by the cooperation activities among JAERI, universities and industries in Japan. The Li2TiO3 pebble fabricated by wet process is a reference material as a tritium breeder, but the stability on high temperature has to be improved for application to DEMO blanket. As one of such the improved materials, TiO2-doped Li2TiO3 pebbles were successfully fabricated and TiO2-doped Li2TiO3 has been studied. For the advanced neutron multiplier, the beryllides that have high melting point and good chemical stability have been studied. Some characterization of Be12Ti was conducted, and it became clear that Be12Ti had lower swelling and tritium inventory than that of beryllium metal. The pebble fabrication study for Be12Ti was also performed and Be12Ti pebbles were successfully fabricated. From these activities, the bright prospect was obtained to realize the DEMO blanket by the application of TiO2-doped Li2TiO3 and beryllides. (author)

  3. Fast breeder reactors in relation to energy requirements. Chapter 2.2

    International Nuclear Information System (INIS)

    It is shown that the world is going to need substantial quantities of energy from new sources early in the next century. Although it may be possible to get a significant amount from solar and geothermal sources, it is far too early to predict how large a contribution they will be able to make. Nuclear power can make a large contribution and it would be wrong to do anything to close this option at this stage. Although there is considerable uncertainty about the precise quantity of commercially exploitable uranium in the world, it is almost certain that breeder reactors will be required. The time-scale is such that utilities throughout the world will need to be able to order breeders in quantity with complete confidence for operation from the mid 1990s onwards. The engineering, safety and logistic problems that are considered in the other chapters of this book are such that, if we are to meet this time-scale, we must press ahead with the utmost urgency. (author)

  4. Reactor fueling of BWR type reactors

    International Nuclear Information System (INIS)

    Purpose: To enable the pattern exchange for control rods during burning in Control Cell Core type BWR reactors. Constitution: A plurality of control cells are divided into a plurality of groups such that the control cells is aparted from each other by way of at least two fuel assemblies other than the control cells with respect to the vertical and lateral directions of the reactor core cross section, as well as they are in adjacent with control cells of other groups with respect to the orthogonal direction. This enables to perform the pattern exchange for the control rods during burning in the control cell core with ease, and the control blade and the story effect harmful to the mechanical soundness of fuels can thus be suppressed. (Moriyama, K.)

  5. Nuclear reactors and fuel cycle

    International Nuclear Information System (INIS)

    The Nuclear Fuel Center (CCN) of IPEN produces nuclear fuel for the continuous operation of the IEA-R1 research reactor of IPEN. The serial production started in 1988, when the first nuclear fuel element was delivered for IEA-R1. In 2011, CCN proudly presents the 100th nuclear fuel element produced. Besides routine production, development of new technologies is also a permanent concern at CCN. In 2005, U3O8 were replaced by U3Si2-based fuels, and the research of U Mo is currently under investigation. Additionally, the Brazilian Multipurpose Research Reactor (RMB), whose project will rely on the CCN for supplying fuel and uranium targets. Evolving from an annual production from 10 to 70 nuclear fuel elements, plus a thousand uranium targets, is a huge and challenging task. To accomplish it, a new and modern Nuclear Fuel Factory is being concluded, and it will provide not only structure for scaling up, but also a safer and greener production. The Nuclear Engineering Center has shown, along several years, expertise in the field of nuclear, energy systems and correlated areas. Due to the experience obtained during decades in research and technological development at Brazilian Nuclear Program, personnel has been trained and started to actively participate in design of the main system that will compose the Brazilian Multipurpose Reactor (RMB) which will make Brazil self-sufficient in production of radiopharmaceuticals. The institution has participated in the monitoring and technical support concerning the safety, licensing and modernization of the research reactors IPEN/MB-01 and IEA-R1. Along the last two decades, numerous specialized services of engineering for the Brazilian nuclear power plants Angra 1 and Angra 2 have been carried out. The contribution in service, research, training, and teaching in addition to the development of many related technologies applied to nuclear engineering and correlated areas enable the institution to fulfill its mission that is to

  6. Testing and qualification of Control and Safety Rod and its drive mechanism of Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Prototype Fast Breeder Reactor (PFBR) has two independent fast acting diverse shutdown systems. The absorber rod of the first system is called Control and Safety Rod (CSR). CSR and its Drive Mechanism (CSRDM) are used for reactor control and for safe shutdown of the reactor by scram action. In view of the safety role, the qualification of CSRDM is one of the important requirements. CSR and CSRDM were qualified in two stages by extensive testing. In the first stage, the critical subassemblies of the mechanism, such as scram release electromagnet, hydraulic dashpot and dynamic seals and CSR subassembly, were tested and qualified individually simulating the operating conditions of the reactor. Experiments were also carried out on sodium vapour deposition in the annular gaps between the stationary and mobile parts of the mechanism. In the second stage, full-scale CSRDM and CSR were subjected to all the integrated functional tests in air, hot argon and subsequently in sodium simulating the operating conditions of the reactor and finally subjected to endurance tests. Since the damage occurring in CSRDM and CSR is mainly due to fatigue cycles during scram actions, the number of test cycles was decided based on the guidelines given in ASME, Section III, Div. 1. The results show that the performance of CSRDM and CSR is satisfactory. Subsequent to the testing in sodium, the assemblies having contact with liquid sodium/sodium vapour were cleaned using CO2 process and the total cleaning process has been established, so that the mechanism can be reused in sodium. The various stages of qualification programmes have raised the confidence level on the performance of the system as a whole for the intended and reliable operation in the reactor.

  7. Evaluation of the Initial Isothermal Physics Measurements at the Fast Flux Test Facility, a Prototypic Liquid Metal Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess

    2010-03-01

    The Fast Flux Test Facility (FFTF) was a 400-MWt, sodium-cooled, low-pressure, high-temperature, fast-neutron flux, nuclear fission reactor plant designed for the irradiation testing of nuclear reactor fuels and materials for the development of liquid metal fast breeder reactors (LMFBRs). The FFTF was fueled with plutonium-uranium mixed oxide (MOX) and reflected by Inconel-600. Westinghouse Hanford Company operated the FFTF as part of the Hanford Engineering Development Laboratory (HEDL) for the U.S. Department of Energy on the Hanford Site near Richland, Washington. Although the FFTF was a testing facility not specifically designed to breed fuel or produce electricity, it did provide valuable information for LMFBR projects and base technology programs in the areas of plant system and component design, component fabrication, prototype testing, and site construction. The major objectives of the FFTF were to provide a strong, disciplined engineering base for the LMFBR program, provide fast flux testing for other U.S. programs, and contribute to the development of a viable self-sustaining competitive U.S. LMFBR industry. During its ten years of operation, the FFTF acted as a national research facility to test advanced nuclear fuels, materials, components, systems, nuclear power plant operating and maintenance procedures, and active and passive reactor safety technologies; it also produced a large number of isotopes for medical and industrial users, generated tritium for the U.S. fusion research program, and participated in cooperative, international research work. Prior to the implementation of the reactor characterization program, a series of isothermal physics measurements were performed; this acceptance testing program consisted of a series of control rod worths, critical rod positions, subcriticality measurements, maximum reactivity addition rates, shutdown margins, excess reactivity, and isothermal temperature coefficient reactivity. The results of these

  8. Clinch River Breeder Reactor Plant Steam Generator Few Tube Test model post-test examination

    International Nuclear Information System (INIS)

    The Steam Generator Few Tube Test (FTT) was part of an extensive testing program carried out in support of the Clinch River Breeder Reactor Plant (CRBRP) steam generator design. The testing of full-length seven-tube evaporator and three-tube superheater models of the CRBRP design was conducted to provide steady-state thermal/hydraulic performance data to full power per tube and to verify the absence of multi-year endurance problems. This paper describes the problems encountered with the mechanical features of the FTT model design which led to premature test termination, and the results of the post-test examination. Conditions of tube bowing and significant tube and tube support gouging was observed. An interpretation of the visual and metallurgical observations is also presented. The CRBRP steam generator has undergone design evaluations to resolve observed deficiences found in the FFTM

  9. Global depletion analysis of Korean helium cooled solid breeder TBM model for demo fusion reactor

    International Nuclear Information System (INIS)

    The Korean HCSB (helium cooled solid breeder) TBM (test blanket module) is proposed with its specific compositions of lithium ceramic, beryllium and graphite in pebble form. In the Korean HCSB TBM, the amount of beryllium is reduced and the reduction is replaced by graphite for a neutron reflector, while tritium breeding ratio (TBR) remains almost unchanged with relatively low Li6 enrichment of ∼40%. However, the previous Korean HCSB was designed based on the LOCAL assumption, in which the surroundings are assumed by the reflective boundary condition. In this research, we establish a simple GLOBAL neutronics model based on demo fusion reactor and perform neutronics analyses including depletion (transmutation) calculation during 100 EFPDs (effective full power days) using the modified MONTEBURNS code.

  10. Recovery of tritium dissolved in sodium at the steam generator of fast breeder reactor

    International Nuclear Information System (INIS)

    The tritium recovery technique in steam generators for fast breeder reactors using the double pipe concept was proposed. The experimental system for developing an effective tritium recovery technique was developed and tritium recovery experiments using Ar gas or Ar gas with 10-10000 ppm oxygen gas were performed using D2 gas instead of tritium gas. It was found that deuterium permeation through two membranes decreased by installing the double pipe concept with Ar gas. By introducing Ar gas with 10000 ppm oxygen gas, the concentration of deuterium permeation through two membranes decreased by more than 1/200, compared with the one pipe concept, indicating that most of the deuterium was scavenged by Ar gas or reacted with oxygen to form a hydroxide. However, most of the hydroxide was trapped at the surface of the membranes because of the short duration of the experiment. (authors)

  11. Development of an ISI Robot for the Fast Breeder Reactor MONJU Primary Heat Transfer System Piping

    International Nuclear Information System (INIS)

    This paper describes the development of a new inspection robot for the In-Service Inspection of the heat transfer system of the Fast Breeder Reactor MONJU. The inspection was carried out using a tire type ultrasonic sensor for volumetric tests at high temperature (atmosphere 55 degree C, Piping Surface 80 degree C) and radiation exposure condition (dose rate 10 mGy/h, piping surface dose rate 15 mGy/h). It was developed an inspection robot using a new tire type for the ultrasonic testing sensor and a new control method. A signal to noise ratio S/N over 2 was obtained during the functional test for a calibration defect with depth 50%t (from the tube wall thickness). (author)

  12. Ultrasonic inspection of liquid-metal fast breeder reactor steam generator duplex tubing

    International Nuclear Information System (INIS)

    Two ultrasonic inspections of the Experimental Breeder Reactor II steam generator duplex tubing have been completed. Inspections performed on one evaporator in 1976 provided baseline data, and a subsequent inspection in 1978 revealed no change in tube condition. With the completion of the 1978 inspection, all available tubes in one evaporator have been inspected. The steam generator contains duplex tubes fabricated from 2 1/4 Cr-1 Mo ferritic steel. Access to the bore (water) side of the tubes was gained through the steam outlet piping. The inspection included a complete volumertic (100% of the tube material) examination, measurement of wall thickness, and evaluation of the condition of the braze bonding the two walls of the tube together. The test equipment was routinely calibrated against a standard containing artificial flaws. Artificial flaws as small as 1.6 mm long x 0.25 mm deep were readily detected

  13. Fission and corrosion product behaviour in liquid metal fast breeder reactors (LMFBRs)

    International Nuclear Information System (INIS)

    It is intended that this review will be useful not only to scientists but also to those concerned with design, day-to-day operation of plant, with liquid metal fast breeder reactors (LMFBRs), safety and decommissioning. Because of this, the review has been widened to include not only the mass transfer behaviour of the various radionuclides in experimental and operating systems, but also the monitoring of the various species, the methods of measurement and the development of methods to control the build-up of the more important long half-life species in operating plants. The information used in the review has been taken from open literature sources to provide an up-to-date presentation of the behaviour of the various isotopes in LMFBRs. 172 refs, 14 figs, 22 tabs

  14. Clinch River Breeder Reactor: an assessment of need for power and regulatory issues

    Energy Technology Data Exchange (ETDEWEB)

    Hamblin, D M; Tepel, R C; Bjornstad, D J; Hill, L J; Cantor, R A; Carroll, P J; Cohn, S M; Hadder, G R; Holcomb, B D; Johnson, K E

    1983-09-01

    The purpose of this report is to present the results of a research effort designed to assist the US Department of Energy in: (1) reviewing the need for power from the Clinch River Breeder Reactor (CRBR) in the Southeastern Electric Reliability Council (SERC) region, not including Florida, and (2) isolating specific regulatory and institutional issues and physical transmission capacities that may constrain the market for CRBR power. A review of existing electric power wheeling arrangements in the Southeast and specific federal and state regulatory obstacles that may affect power sales from the CRBR was undertaken. This review was a contributing factor to a decision to target the service territory to SERC-less Florida.

  15. Numerical simulation of sodium pool fires in liquid metal-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    In Liquid Metal-Cooled Fast Breeder Reactor (LMFBR), the leakage of sodium can result in sodium fires. Due to sodium's high chemical reactivity in contact with air and water, sodium fires will lead to an immediate increase of the air temperature and pressure in the containment. This will harm the integrity of the containment. In order to estimate and foresee the sequence of this accident, or to prevent the accident and alleviate the influence of the accident, it is necessary to develop programs to analyze such sodium fire accidents. Based on the work of predecessors, flame sheet model is produced and used to analyze sodium pool fire accidents. Combustion model and heat transfer model are included and expatiated. And the comparison between the analytical and experimental results shows the program is creditable and reasonable. This program is more realistic to simulate the sodium pool fire accidents and can be used for nuclear safety judgement. (authors)

  16. Constituent migration model for fast reactor U-Pu-Zr metallic fuel

    International Nuclear Information System (INIS)

    The metallic fuel behavior of U-Pu-Zr has been actively investigated at Argonne National Laboratory. The Central Research Institute of the Electric Power Industry has been developing a metallic fuel liquid-metal fast breeder reactor concept jointly with several Japanese organizations. One of the development activities resulted in the fuel performance code SESAME. Recently, joint efforts have included the development of more mechanistic models for metallic fuel behavior. One of the models is for the constituent migration behavior in U-Pu-Zr fuels. For a simulation of constituent migration in the Experimental Breeder Reactor 2 (EBR-2) test pins ND30 and ND35, a quasi-binary U-Zr-(11.5 at.% Pu) phase diagram is produced using the U-Pu-Zr ternary phase diagram of O'Boyle and Dwight

  17. Nuclear reactors. To breed or not to breed. A Pugwash debate on fast breeder reactors held at the Royal Society, London, on 28 September 1976 under the chairmanship of Sir Alec Merrison

    International Nuclear Information System (INIS)

    The debate which is reported was timed to coincide with the publication of the Report of the (UK) Royal Commission on Environmental Pollution: 'Nuclear Power and Environment'. The volume comprises an introductory section, a report of an address by the Chairman of the Royal Commission and invited papers on fast breeder reactors in relation to energy requirements, on the safety of a commercial fast reactor, on processing and reprocessing of fuel, on radioactive waste management, and on diversion of plutonium and proliferation of nuclear weapons. An edited version of the discussion is presented under the following heads: comments on the report of the Royal Commission; projections of future energy requirement; thermal pollution; safety and insurance of reactors; reprocessing of fuel; storage and disposal of wastes; energy from fusion; utilization of coal; and proliferation of weapons and diversion of plutonium. The six invited papers are considered to be within INIS scope and separate abstracts have been prepared. (U.K.)

  18. Development of high temperature fission counter-chamber(FC)S for a top entry loop type fast breeder reactor

    International Nuclear Information System (INIS)

    Prototype high temperature fission counter-chambers have been made as neutron detectors for installation in the reactor vessel of the 600MWe-class top entry loop type fast breeder reactor. Using these prototypes as samples, a high-temperature endurance test has been conducted. The validity of the prototypes has been established by the test results, which show that the prototypes nearly satisfy the design performance. (author)

  19. Recommended practices in elevated temperature design: A compendium of breeder reactor experiences (1970-1986): An overview

    International Nuclear Information System (INIS)

    Significant experiences have been accumulated in the establishment of design methods and criteria applicable to the design of Liquid Metal Fast Breeder Reactor (LMFBR) components. The Subcommittee of the Elevated Temperature Design under the Pressure Vessel Research Council (PVRC) has undertaken to collect, on an international basis, design experience gained, and the lessons learned, to provide guidelines for next generation advanced reactor designs. This paper shall present an overview and describe the highlights of the work

  20. Nuclear reactors and fuel cycle

    International Nuclear Information System (INIS)

    The Center for Nuclear Engineering has shown expertise in the field of nuclear and energy systems ad correlated areas. Due to the experience obtained over decades in research and technological development at Brazilian Nuclear Program personnel has been trained and started to actively participate in the design of the main system that will compose the Brazilian Multipurpose Reactor (RMB) which will make Brazil self-sufficient in the production of radiopharmaceuticals. The institution has participated in the monitoring and technical support concerning the safety, licensing and modernization of the research reactors IPEN/MB-01 and IEA-R1. The Nuclear Fuel Center is responsible for the production of the nuclear fuel necessary for the continuous operation of the IEA-R1 research reactor. Development of new fuel technologies is also a permanent concern

  1. Modeling and Simulation of Operator Training Simulator for Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Often the challenge faced by the Nuclear Power Industry is the availability of well trained human resource for efficient power plant monitoring and control. Safety of the plant purely depends upon the plant knowledge acquired, understanding of plant dynamics and the skills possessed by the operators through systematic training. Generally the operators are given class room and field training before deploying them in the operation crew. But, for handling emergency and abnormal conditions, the formal class room training and field training have proved to be inadequate according to the expert estimates. The state- of- art computer based operator training simulators covering the full spectrum of the plant have become an essential element in bridging the gap between the inadequacy and efficiency. Gradually the training simulators are getting embedded in the operator training programme and started playing a crucial role in enhancing the ability of the operators.This paper discusses about the operator training simulator called KALBR- SIM i.e. Kalpakkam Breeder Simulator that has been built at IGCAR for training the Prototype Fast Breeder Reactor (PFBR) operators. It is a Full Scope Replica Operator Training Simulator built to replicate PFBR. The scope of the paper covers the basic modules necessary for building each process model of the simulator, design and development of the reactor sub systems like Neutronics, Primary Sodium, Secondary Sodium, Decay Heat Removal, Steam Water, Electrical systems and the associated logics and controls. It is followed by a detailed discussion on replication aspects of Simulator Control Room and its advantages, the Hardware Architecture, Instruction Station facility and loading of scenarios. It further elaborates on Steady State and Bench Mark Transients tests conducted on the Operator Training Simulator like One primary sodium pump trip, one primary pump seizure, Primary pipe rupture, one boiler feed pump trip and Station Black Out. (author)

  2. Acoustic monitoring of sodium boiling in a liquid metal fast breeder reactor from autoregressive models

    Energy Technology Data Exchange (ETDEWEB)

    Geraldo, Issa Cherif [Laboratoire d’Automatique, Génie Informatique et Signal (LAGIS UMR CNRS 8219), Université Lille 1, Sciences et technologies, Avenue Paul Langevin, BP 48, 59651 Villeneuve d’Ascq CEDEX (France); Bose, Tanmoy [Indian Institute of Technology Kharagpur, Kharagpur 721302, West Bengal (India); Pekpe, Komi Midzodzi, E-mail: midzodzi.pekpe@univ-lille1.fr [Laboratoire d’Automatique, Génie Informatique et Signal (LAGIS UMR CNRS 8219), Université Lille 1, Sciences et technologies, Avenue Paul Langevin, BP 48, 59651 Villeneuve d’Ascq CEDEX (France); Cassar, Jean-Philippe [Laboratoire d’Automatique, Génie Informatique et Signal (LAGIS UMR CNRS 8219), Université Lille 1, Sciences et technologies, Avenue Paul Langevin, BP 48, 59651 Villeneuve d’Ascq CEDEX (France); Mohanty, A.R. [Indian Institute of Technology Kharagpur, Kharagpur 721302, West Bengal (India); Paumel, Kévin [CEA, DEN, Nuclear Technology Department, F-13108 Saint-Paul-lez-Durance (France)

    2014-10-15

    Highlights: • The work deals with sodium boiling detection in a liquid metal fast breeder reactor. • The authors choose to use acoustic data instead of thermal data. • The method is designed to not to be disturbed by the environment noises. • A real time boiling detection methods are proposed in the paper. - Abstract: This paper deals with acoustic monitoring of sodium boiling in a liquid metal fast breeder reactor (LMFBR) based on auto regressive (AR) models which have low computational complexities. Some authors have used AR models for sodium boiling or sodium–water reaction detection. These works are based on the characterization of the difference between fault free condition and current functioning of the system. However, even in absence of faults, it is possible to observe a change in the AR models due to the change of operating mode of the LMFBR. This sets up the delicate problem of how to distinguish a change in operating mode in absence of faults and a change due to presence of faults. In this paper we propose a new approach for boiling detection based on the estimation of AR models on sliding windows. Afterwards, classification of the models into boiling or non-boiling models is made by comparing their coefficients by two statistical methods, multiple linear regression (LR) and support vectors machines (SVM). The proposed approach takes into account operating mode information in order to avoid false alarms. Experimental data include non-boiling background noise data collected from Phenix power plant (France) and provided by the CEA (Commissariat à l’Energie Atomique et aux énergies alternatives, France) and boiling condition data generated in laboratory. High boiling detection rates as well as low false alarms rates obtained on these experimental data show that the proposed method is efficient for boiling detection. Most importantly, it shows that the boiling phenomenon introduces a disturbance into the AR models that can be clearly detected.

  3. Development of a nuclear information system for the MONJU Fast Breeder Reactor

    International Nuclear Information System (INIS)

    At the MONJU Fast Breeder Reactor information is collected on a continuous basis. This information consists of measured data, design data, simulations data, maintenance data etc. which may be in any kind of electronic form, i.e. text documents, code input files, reports or even scanned documents. The amount and nature of these data has introduced the need for a software system, which will provide an efficient infrastructure for the maintenance of and operations on the data. Thus a Nuclear Information System for the MONJU Fast Breeder Reactor is under development. The system consists of remote databases hosting the information along with clients for handling them, remote clients providing the users with an interface and a local server for handling the client requests and the communication between the database and user clients. The system is composed of independent server, database and user modules, which communicates using the RMI-IIOP (Remote Method Invocation - Internet InterORB Protocol) technology. The RMI-IIOP is a CORBA (Common Object Request Broker Architecture) compliant subset of the RMI thereby facilitating the possibility of implementing the database and user modules in any kind of programming language and on any kind of operating system by providing a standard, platform independent communications interface. The user interface consists of dynamic HTML web pages which instantiates servlets in the user module when the user submits queries. The database module consists of controllers for handling the communication with the user module and database drivers for handling the connections with the databases. In this paper the overall system design and schemes for data flow and remote method invocations are presented and the requirements imposed on the system are discussed. (author)

  4. Acoustic monitoring of sodium boiling in a liquid metal fast breeder reactor from autoregressive models

    International Nuclear Information System (INIS)

    Highlights: • The work deals with sodium boiling detection in a liquid metal fast breeder reactor. • The authors choose to use acoustic data instead of thermal data. • The method is designed to not to be disturbed by the environment noises. • A real time boiling detection methods are proposed in the paper. - Abstract: This paper deals with acoustic monitoring of sodium boiling in a liquid metal fast breeder reactor (LMFBR) based on auto regressive (AR) models which have low computational complexities. Some authors have used AR models for sodium boiling or sodium–water reaction detection. These works are based on the characterization of the difference between fault free condition and current functioning of the system. However, even in absence of faults, it is possible to observe a change in the AR models due to the change of operating mode of the LMFBR. This sets up the delicate problem of how to distinguish a change in operating mode in absence of faults and a change due to presence of faults. In this paper we propose a new approach for boiling detection based on the estimation of AR models on sliding windows. Afterwards, classification of the models into boiling or non-boiling models is made by comparing their coefficients by two statistical methods, multiple linear regression (LR) and support vectors machines (SVM). The proposed approach takes into account operating mode information in order to avoid false alarms. Experimental data include non-boiling background noise data collected from Phenix power plant (France) and provided by the CEA (Commissariat à l’Energie Atomique et aux énergies alternatives, France) and boiling condition data generated in laboratory. High boiling detection rates as well as low false alarms rates obtained on these experimental data show that the proposed method is efficient for boiling detection. Most importantly, it shows that the boiling phenomenon introduces a disturbance into the AR models that can be clearly detected

  5. Fuel for advanced CANDU reactors

    International Nuclear Information System (INIS)

    The CANDU reactor system has proven itself to be a world leader in terms of station availability and low total unit energy cost. In 1985 for example, four of the top ten reactor units in the world were CANDU reactors operating in South Korea and Canada. This excellent operating record requires an equivalent performance record of the low-cost, natural uranium fuel. Future CANDU reactors will be an evolution of the present design. Engineering work is under way to refine the existing CANDU 600 and to incorporate state-of-the-art technology, reducing the capital cost and construction schedule. In addition, a smaller CANDU 300 plant has been designed using proven CANDU 600 technology and components but with an innovative new plant layout that makes it cost competitive with coal fired plants. For the long term, work on advanced fuel cycles and major system improvements is underway ensuring that CANDU plants will stay competitive well into the next century

  6. Nuclear reactor fuel element

    International Nuclear Information System (INIS)

    The grid-shaped spacer for PWR fuel elements consists of flat, upright metal bars at right angles to the fuel rods. In one corner of a grid mesh it has a spring with two end parts for the fuel rod. The cut-outs for the end parts start from an end edge of the metal bar parallel to the fuel rods. The transverse metal bar is one of four outer metal bars. Both end parts of the spring have an extension parallel to this outer metal arm, which grips a grid mesh adjacent to this grid mesh at the side in one corner of the spacer and forms an end part of a spring for the fuel rod there on the inside of the outer metal bar. (HP)

  7. Criticality of spent reactor fuel

    International Nuclear Information System (INIS)

    The storage capacity of spent reactor fuel pools can be greatly increased by consolidation. In this process, the fuel rods are removed from reactor fuel assemblies and are stored in close-packed arrays in a canister or skeleton. An earlier study examined criticality consideration for consolidation of Westinghouse fuel, assumed to be fresh, in canisters at the Millstone-2 spent-fuel pool and in the General Electric IF-300 shipping cask. The conclusions were that the fuel rods in the canister are so deficient in water that they are adequately subcritical, both in normal and in off-normal conditions. One potential accident, the water spill event, remained unresolved in the earlier study. A methodology is developed here for spent-fuel criticality and is applied to the water spill event. The methodology utilizes LEOPARD to compute few-group cross sections for the diffusion code PDQ7, which then is used to compute reactivity. These codes give results for fresh fuel that are in good agreement with KENO IV-NITAWL Monte Carlo results, which themselves are in good agreement with continuous energy Monte Carlo calculations. These methodologies are in reasonable agreement with critical measurements for undepleted fuel

  8. Alternate-fuel reactor studies

    International Nuclear Information System (INIS)

    A number of studies related to improvements and/or greater understanding of alternate-fueled reactors is presented. These studies cover the areas of non-Maxwellian distributions, materials and lifetime analysis, a 3He-breeding blanket, tritium-rich startup effects, high field magnet support, and reactor operation spanning the range from full D-T operation to operation with no tritium breeding

  9. FPGA based pump speed measurement system for prototype fast breeder reactor

    International Nuclear Information System (INIS)

    In Prototype Fast Breeder Reactor (PFBR), the heat generated in the reactor core by nuclear fission is extracted by circulating liquid sodium through it using two Primary Sodium Pumps (PSP). The FPGA-based PSP Speed measurement system is a safety critical system provided to protect the reactor in the event of a PSP seizure. The function of the system is to measure the PSP speed and initiate safety action if there is a reduction in speed below a specified set point. Variable Reluctance Sensor (VRS) is used for measuring the PSP speed. This sensor outputs a voltage signal, whose amplitude and frequency are proportional to the pump speed. The frequency of the sensor signal is measured, translated to the pump speed, and compared with user-specified set points for generating the required alarm and trip (safety action) signals. This paper explains the system requirements, system architecture, implementation and qualification tests carried out on the system. Since the timing requirements on the system are stringent, a pipelined architecture is used for improving the system response time, which is detailed in this paper. Since the system is safety critical, various safety and failsafe features are incorporated in the system which are also explained. (author)

  10. Dynamic modeling of steam water system of prototype fast breeder reactor using RELAP code

    International Nuclear Information System (INIS)

    Highlights: • Dynamic modeling of steam water system of an LMBFR using RELAP5 code. • Analysis of events managed by power setback procedure in the plant. • Selection of parameters and their thresholds for the power setback procedure. - Abstract: The safety of a fast breeder reactor based nuclear power plant with an intermediate coolant loop between the primary and tertiary circuit, depends on the correct functioning of actions initiated by the balance-of-plant (tertiary) systems. Comfortable time will be available for such actions to ensure reactor safety. Perturbations in the Balance-of-Plant (BoP) influence the transient sequence of safety-relevant parameters of the plant in a benign manner. However, for complete and realistic prediction of transient behavior of the whole plant, dynamic models for BoP systems are required to be developed. This paper describes modeling of BoP system of PFBR using RELAP5/MOD 3.4 code. Some of the important transients in the BoP system, which are managed by reactor power setback procedure have been analysed using this code to verify the effectiveness of the procedure adopted

  11. Method for separating neptunium from irradiated nuclear fuel and/or breeder materials

    International Nuclear Information System (INIS)

    The proposed method is an addition to the method described in patent 1905519 for separating uranium and plutonium compounds from irradiated fuel or breeder elements by liquid-liquid extraction, where the plutonium from the uranium-retaining organic phase is directly reduced during the extraction process and separated of with the aqueous phase. It is now proposed to electrolytically reduce the neptunium together with the plutonium in one process step. Both are thus practically completely transfered to the aqueous phase and thus separated from uranium. (orig./PW)

  12. Power flattening of an inertial fusion energy breeder with mixed ThO2-UO2 fuel

    International Nuclear Information System (INIS)

    Neutronic performance of a blanket-driven ICF (inertial confinement fusion) neutron and based on SiCf/SiC composite material has been investigated for fissile fuel breeding and a flat fission power density. The blanket is fueled with ThO2 and UO2 mixed by various mixing methods for achieving a flat fission power density, and cooled with different coolants, natural lithium, (LiF)2BeF2, Li17Pb83 and 4He for the nuclear heat transfer. MCNP4B code is used for calculations of neutronic data per DT fusion neutron. M (fusion energy multiplication rate) increases to ∼2.6 in the 4He coolant blanket. On the other hand, this value is ∼2.00 in the Li17Pb83 coolant blanket because of more capture reactions than fission reactions in this blanket. Therefore, the investigated reactor can produce substantial electricity in situ. Total fissile fuel breading ratio (FFBR) (232Thγ+238Uγ) is increased from ∼0.16 to ∼0.37 by using coolants having high neutron multiplier cross-section. Values of TBR (tritium breeding ratio) being one of the main neutronic parameters for a fusion reactor are greater than 1.05 in all case of mixed fuels and type of coolants. The highest and the lowest TBR values are ∼1.34 and ∼1.07 in the natural Li and the 4He coolant blanket, respectively. Consequently, the breeder hybrid reactor is self-sufficient in the tritium required for the DT fusion driver. Γ (peek-to-average fission power density ratio) of the blanket is reduced to ∼1.1. Furthermore, quasi-constant fission power density profile is obtained in FFB (fissile fuel breeding) zone for case of PMF (parabolic mixed fuel) and all type of coolants. Values of neutron leakage out of the blanket in all case of mixed fuels and type of coolants are quite low due to SiC reflector and B4C shielding. The maximum neutron leakage is only ∼0.025

  13. Development of fluorocarbon rubber for backup seals of sodium cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Highlights: → Negligible chemical degradation of seal compound during ageing (in unstrained state) in air at 140/170/200 oC for 32 weeks. → Cross-link exchange, Joule-Gough effect and ionic interaction during ageing in unstrained state. → Enhanced physical/chemical degradation of compound during ageing under strain. → Capability of compound to withstand heat, radiation, air and mechanical load in reactor for 10 years. → Negligible chemical dose rate effect and gas evolution from compound during seal operation. -- Abstract: The development of a fluorohydrocarbon rubber compound for static backup seals of 500 MWe, Prototype Fast Breeder Reactor (PFBR) is depicted. Variations of a previously developed Viton A-401C based formulation were subjected to processability tests, accelerated heat ageing in air, mechanical characterization and production trials. Finite element analysis and literature data extrapolation were combined with long term ageing to ascertain the life (minimum 10 years) of chosen formulation in reactor under synergistic influences of 110 oC, 23 mGy/h (γ dose rate) and air considering postulated accidental conditions. Validation of test seals and quality assessment indicate that composition and properties of the validated laboratory compound has been translated effectively to the reactor seals, installed recently in PFBR. The tensile and hardness specimens indicated negligible degradation and exceptional thermo-oxidative stability of the seal compound during ageing (32 weeks at 140/170/200 oC) even though interesting manifestations of cross-link exchange and ionic interactions were observed. Compression set results, showing definite trends of change under ageing and stain, were used in Arrhenius and Williams Landel Ferry equations for realistic life prediction. The development provides a foundation to simplify and standardize the design, development and operation of major elastomeric sealing applications of Indian nuclear reactors based on a

  14. Nuclear reactor fuel element

    International Nuclear Information System (INIS)

    The fuel element for a BWR known from the patent application DE 2824265 is developed so that the screw only breaks on the expansion shank with reduced diameter if the expansion forces are too great. (HP)

  15. Analysis of thorium/U-233 lattices and cores in a breeder/burner heavy water reactor

    International Nuclear Information System (INIS)

    Due to the inevitable dwindling of uranium resources, advanced fuel cycles in the current generation of reactors stand to be of great benefit in the future. Heavy water moderated reactors have much potential to make use of thorium, a currently unexploited resource. Core fuelling configurations of a Heavy Water Reactor based on the self-sufficient thorium fuel cycle were simulated using the DRAGON and DONJON reactor physics codes. Three heterogeneously fuelled reactors and one homogeneously fuelled reactor were studied. (author)

  16. Modelling fuel behaviour in a reactor park using fuel cycle kinetics

    International Nuclear Information System (INIS)

    In this thesis the theory of fuel cycle kinetics is re-examined. The fuel cycle kinetics theory is a powerful tool to describe the time-dependent fuel behaviour of large populations of nuclear reactors. The fuel cycle kinetics theory is based on the point kinetics theory and the principles of a reactor park. The point kinetics theory is a simplification of the space-, energy-and time-dependent diffusion balance equation to only a time-dependent equation. A reactor park is the description of the interconnections between a population of nuclear reactors with various designs. In the fuel cycle kinetics theory the point kinetics theory is used as a model to simplify space- energy- and time-dependent burn-up equations of the reactors in a reactor park to a set of only time-dependent equations, one for every reactor type. The fuel cycle kinetics theory is verified by means of a number of test cases. In the first test case the same symbiotic system is used as was used by Maudlin. There is no difference between the two obtained results. The second test case is that of only Fast Breeder Reactor, FBR, deployment. Here the result of the fuel cycle kinetics equation is checked against the result obtained from TRITON. TRITON is a module of the SCALE code system that is used for depletion analysis of 3-D reactor models. With the use of the pseudo-initial condition the results of the fuel cycle kinetics and TRITON calculations are almost identical. The pseudo-initial condition is a correction on the initial condition to adjust for neglecting the time dependency of the parameters in the fuel cycle kinetics equations. In the third case a symbiotic system of FBRs and Pressurised Water Reactors, PWRs, is researched. There is only a small difference in the asymptotic growth between the fuel cycle kinetics results and the TRITON results. In the last test case the same system of FBRs and PWRs is used to investigate two demanded asymptotic growths obtained from the upper and lower

  17. The role of fast breeder reactors in the future energy economy

    International Nuclear Information System (INIS)

    In this contribution, the reactor design and performance characteristics of a wide range of design concepts are documented. Since the technical feasibility of various design options and associated breeding performance depend strongly on the status of fuels development, primary focus is given to review of oxide, carbide and metal fuels in terms of current technology status, irradiation performance, associated key issues and main R and D requirements to resolve the issues

  18. Analysis of unprotected transients with control and safety rod drive mechanism expansion feedback in a medium sized oxide fuelled fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sathiyasheela, T., E-mail: sheela@igcar.gov.in; Natesan, K.; Srinivasan, G.S.; Devan, K.; Puthiyavinayagam, P.

    2015-09-15

    Highlights: • Possibilities of enhancing safety under ULOF and UTOP accidents. • CSRDM expansion feedbacks under unprotected transients. • CSRDM expansion feedback enhances the safety of fast reactors. • CSRDM expansion feedbacks ensuring enough time for initiating safety actions. - Abstract: Possibilities of enhancing core safety under unprotected loss of flow (ULOF) and unprotected transient over power (UTOP) accidents with control and safety rod drive mechanism (CSRDM) expansion feedbacks are explored in a medium sized oxide fuelled fast breeder reactor. This feedback is expected to take the reactor to a safe shutdown under ULOF and to an another steady state under UTOP where there is no significant fuel melting. Under ULOF, with CSRDM feedback net reactivity was maintained negative throughout the transient (up to 2000 s) and the power dropped to a level of heat removal capacity of decay heat removal system based on natural circulation. Similarly, under UTOP with the above feedback reactor power goes to a lower peak value. The fuel temperature is just touching the melting temperature and the melt fraction does not cross 5%. With CSRDM expansion feedbacks both ULOF and UTOP transients prolong beyond 2000 s. It ensures, availability of time for initiating any safety actions against the transients, and thus it helps to preclude core disruptive accidents (CDA) in a medium sized oxide fuelled reactors.Classification: L. safety and risk analysis.

  19. Transuranic material recovery in the Integral Fast Reactor fuel cycle demonstration

    International Nuclear Information System (INIS)

    The Integral Fast Reactor is an innovative liquid metal reactor concept that is being developed by Argonne National Laboratory. It takes advantage of the properties of metallic fuel and liquid metal cooling to offer significant improvements in reactor safety, operation, fuel cycle economics, environmental protection, and safeguards. The plans for demonstrating the IFR fuel cycle, including its waste processing options, by processing irradiated fuel from the Experimental Breeder Reactor-II fuel in its associated Fuel Cycle Facility have been developed for the first refining series. This series has been designed to provide the data needed for the further development of the IFR program. An important piece of the data needed is the recovery of TRU material during the reprocessing and waste operations

  20. A US perspective on fast reactor fuel fabrication technology and experience part I: metal fuels and assembly design

    International Nuclear Information System (INIS)

    This paper is part I of a review focusing on the United States experience with metallic fast reactor fuel fabrication and assembly design for the Experimental Breeder Reactor-II (EBR-II) and the Fast Flux Test Facility (FFTF). Experience with metal fuel fabrication in the United States is extensive, including over 60 years of research conducted by the government, national laboratories, industry, and academia. This experience has culminated in a considerable amount of research that resulted in significant improvements to the technologies employed to fabricate metallic fast reactor fuel. This part of the review documents the current state of fuel fabrication technologies for metallic fuels, some of the challenges faced by previous researchers, and how these were overcome. Knowledge gained from reviewing previous investigations will aid both researchers and policy makers in forming future decisions relating to nuclear fuel fabrication technologies.

  1. Advanced automation concepts applied to Experimental Breeder Reactor-II startup

    International Nuclear Information System (INIS)

    The major objective of this work is to demonstrate through simulations that advanced liquid-metal reactor plants can be operated from low power by computer control. Development of an automatic control system with this objective will help resolve specific issues and provide proof through demonstration that automatic control for plant startup is feasible. This paper presents an advanced control system design for startup of the Experimental Breeder Reactor-2 (EBR-2) located at Idaho Falls, Idaho. The design incorporates recent methods in nonlinear control with advanced diagnostics techniques such as neural networks to form an integrated architecture. The preliminary evaluations are obtained in a simulated environment by a low-order, valid nonlinear model. Within the framework of phase 1 research, the design includes an inverse dynamics controller, a fuzzy controller, and an artificial neural network controller. These three nonlinear control modules are designed to follow the EBR-2 startup trajectories in a multi-input/output regime. They are coordinated by a supervisory routine to yield a fault-tolerant, parallel operation. The control system operates in three modes: manual, semiautomatic, and fully automatic control. The simulation results of the EBR-2 startup transients proved the effectiveness of the advanced concepts. The work presented in this paper is a preliminary feasibility analysis and does not constitute a final design of an automated startup control system for EBR-2. 14 refs., 43 figs

  2. Real Time Computer for Plugging Indicator Control of Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Prototype Fast Breeder Reactor (PFBR) is in the advanced stage of construction at Kalpakkam, India. Liquid sodium is used as coolant to transfer the heat produced in the reactor core to steam water circuit. Impurities present in the sodium are removed using purification circuit. Plugging indicator is a device used to measure the purity of the sodium. Versa Module Europa bus based Real Time Computer (RTC) system is used for plugging indicator control. Hot standby architecture consisting of dual redundant RTC system with switch over logic system is the configuration adopted to achieve fault tolerance. Plugging indicator can be controlled in two modes namely continuous and discontinuous mode. Software based Proportional-Integral-Derivative (PID) algorithms are developed for plugging indicator control wherein the set point changes dynamically for every scan interval of the RTC system. Set points and PID constants are kept as configurable in runtime in order to control the process in very efficient manner, which calls for reliable communication between RTC system and control station, hence TCP/IP protocol is adopted. Performance of the RTC system for plugging indicator control was thoroughly studied in the laboratory by simulating the inputs and monitored the control outputs. The control outputs were also monitored for different PID constants. Continuous and discontinuous mode plots were generated. (authors)

  3. NDT services of Prototype Fast Breeder Reactor PFBR - a contribution from Blue Star Ltd towards growth

    International Nuclear Information System (INIS)

    Prototype fast breeder reactor is 500MWe sodium cooled reactor of pool type design. Three classes of steels namely austenitic stainless steel (304LN and 316LN), Ferritic steel (Modified 9Cr 1 Mo) and Carbon steel materials (A48P2) were used for manufacturing of different nuclear steam supply system (NSSS) components. The components are in different product forms such as castings, forgings, plates, rounds, Hollow bars, Seamless tubes and pipes for fitness-for- purpose applications. Due to the criticality of the design, stringent quality control measures were to be adopted for component integrity. The collaboration of Blue Star Ltd with major players of fabrication such as M/s. Larsen and Toubro Ltd (main and safety vessels, sodium tanks), Kirloskar Brothers (primary sodium pump forgings), Walchand Nagar Industries (grid plate assembly), Bharat Heavy Electricals (inner vessel and thermal baffle), MTAR Technologies (grid plate sleeves) and Omplas Systems (colmonoy hard facing of various core components) had witnessed many challenges in achieving the required quality. Different code based (RCCMR and ASME) approaches for acceptance criteria had to be adopted for evaluation of the components that are large in dimensions and quantities. This paper discusses the knowledge gained through different procedural developments for these critical components and provided suggestion of remedial measures in achieving the required quality. (author)

  4. Design and fabrication of sodium test facility for fast breeder reactor

    International Nuclear Information System (INIS)

    The purpose of the promotion policy for energy research and development base construction plan (priority facility) of the Japanese government in FY2009 is 'to construct in Tsuruga City the research and development base for plant operation technology for the practical use of fast breeder reactor where researchers in and out of Japan gather, and to contribute to the development and revitalization of the region as the base with international characteristics.' In conformity to this purpose, the Japan Atomic Energy Agency built 'sodium engineering research facilities' in Tsuruga. This paper describes the design, fabrication, and installation of interior equipment that were carried out by Kawasaki Heavy Industries. 'Sodium engineering research facilities' are the test and research facilities to conduct research and development related to sodium, while reflecting the experiences of operation and maintenance of 'Monju,' which aims at the commercialization of fast reactor. The facilities specialize in the handling technology of sodium to meet the needs in and out of Japan, and were completed in June 2015. The facilities consist of six units including tank-loop test equipment, mini-loop test equipment, sodium purification and supply equipment, etc. For the tank-loop test equipment, a sodium transfer test of about 5.5 tons, and a subsequent comprehensive function test using sodium are scheduled. (A.O.)

  5. Design, implementation and cost-benefit analysis of a dynamic testing program in the Experimental Breeder Reactor-II

    International Nuclear Information System (INIS)

    Dynamic tests have been performed for many years in commercial pressurized and boiling water reactors. The purpose of this study was to evaluate the technological and economical feasibility of extending the current light water reactor testing procedures to both present and future liquid metal fast breeder reactors. A 38 node linearized, lumped parameter, EBR-II system model was developed. This model was analyzed to obtain the predicted system time and frequency response for reactivity perturbations, intermediate heat exchanger secondary inlet sodium temperature perturbation frequency response, and various system nodal frequency response sensitivities

  6. Report on the shearing, dissolution and analysis of GRIP-II rod 79-453 (validation rod); Light Water Breeder Reactor proof-of-breeding analytical support project

    International Nuclear Information System (INIS)

    This report covers the processing and analysis of the fuel-bearing section (M-5138) of an irradiated experimental Light Water Breeder Reactor fuel rod, GRIP-II rod No. 79-453; this section has been designated the Validation Rod. Process steps included precision shearing of the rod into eight comminuted segments, dissolution of the segments, and chemical and radiometric analyses of the resulting solutions. The shearing and dissolution were carried out fully remotely in an existing pilot-scale facility installed in a shielded cell. Data are provided on physical parameters of the rod section and segments, uranium assays and isotopic abundances, and selected fission products. An error analysis of the individual measurements and analyses is included

  7. Analysis of the conceptual shielding design for the upflow Gas-Cooled Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Slater, C.O.; Reed, D.A.; Cramer, S.N.; Emmett, M.B.; Tomlinson, E.T.

    1981-01-01

    Conceptual Shielding Configuration III for the Gas-Cooled Fast Breeder Reactor (GCFR) was analyzed by performing global calculations of neutron and gamma-ray fluences and correcting the results as appropriate with bias factors from localized calculations. Included among the localized calculations were the radial and axial cell streaming calculations, plus extensive preliminary calculations and three final confirmation calculations of the plenum flow-through shields. The global calculations were performed on the GCFR mid-level and the lower and upper plenum regions. Calculated activities were examined with respect to the design constraint, if any, imposed on the particular activity. The spatial distributions of several activities of interest were examined with the aid of isoplots (i.e., symbols are used to describe a surface on which the activity level is everywhere the same). In general the results showed that most activities were below the respective design constraints. Only the total neutron fluence in the core barrel appeared to be marginal with the present reactor design. Since similar results were obtained for an earlier design, it has been proposed that the core barrel be cooled with inlet plenum gas to maintain it at a temperature low enough that it can withstand a higher fluence limit. Radiation levels in the prestressed concrete reactor vessel (PCRV) and liner appeared to be sufficiently below the design constraint that expected results from the Radial Shield Heterogeneity Experiment should not force any levels above the design constraint. A list was also made of a number of issues which should be examined before completion of the final shielding design.

  8. Analysis of the conceptual shielding design for the upflow Gas-Cooled Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Conceptual Shielding Configuration III for the Gas-Cooled Fast Breeder Reactor (GCFR) was analyzed by performing global calculations of neutron and gamma-ray fluences and correcting the results as appropriate with bias factors from localized calculations. Included among the localized calculations were the radial and axial cell streaming calculations, plus extensive preliminary calculations and three final confirmation calculations of the plenum flow-through shields. The global calculations were performed on the GCFR mid-level and the lower and upper plenum regions. Calculated activities were examined with respect to the design constraint, if any, imposed on the particular activity. The spatial distributions of several activities of interest were examined with the aid of isoplots (i.e., symbols are used to describe a surface on which the activity level is everywhere the same). In general the results showed that most activities were below the respective design constraints. Only the total neutron fluence in the core barrel appeared to be marginal with the present reactor design. Since similar results were obtained for an earlier design, it has been proposed that the core barrel be cooled with inlet plenum gas to maintain it at a temperature low enough that it can withstand a higher fluence limit. Radiation levels in the prestressed concrete reactor vessel (PCRV) and liner appeared to be sufficiently below the design constraint that expected results from the Radial Shield Heterogeneity Experiment should not force any levels above the design constraint. A list was also made of a number of issues which should be examined before completion of the final shielding design

  9. Reliability analysis of safety grade decay heat removal system of Indian prototype fast breeder reactor

    International Nuclear Information System (INIS)

    The 500 MW Indian pool type Prototype Fast Breeder Reactor (PFBR), is provided with two independent and diverse Decay Heat Removal (DHR) systems viz., Operating Grade Decay Heat Removal System (OGDHRS) and Safety Grade Decay Heat Removal System (SGDHRS). OGDHRS utilizes the secondary sodium loops and Steam-Water System with special decay heat removal condensers for DHR function. The unreliability of this system is of the order of 0.1-0.01. The safety requirements of the present generation of fast reactors are very high, and specifically for DHR function the failure frequency should be less than ∼1E-7/ry. Therefore, a passive SGDHR system using four completely independent thermo-siphon loops in natural convection mode is provided to ensure adequate core cooling for all Design Basis Events. The very high reliability requirement for DHR function is achieved mainly with the help of SGDHRS. This paper presents the reliability analysis of SGDHR system. Analysis is performed by Fault Tree method using 'CRAFT' software developed at Indira Gandhi Centre for Atomic Research. This software has special features for compact representation and CCF analysis of high redundancy safety systems encountered in nuclear reactors. Common Cause Failures (CCF) are evaluated by β factor method. The reliability target for SGDHRS arrived from DHR reliability requirement and the ultimate number of demands per year (7/y) on SGDHRS is that the failure frequency should be ≤1.4E-8/de. Since it is found from the analysis that the unreliability of SGDHRS with identical loops is 5.2E-6/de and dominated by leak rates of components like AHX, DHX and sodium dump and isolation valves, options with diversity measures in important components were studied. The failure probability of SGDHRS for a design consisting of 2 types of diverse loops (Diverse AHX, DHX and sodium dump and isolation valves) is 2.1E-8/de, which practically meets the reliability requirement

  10. Web-enabled work permit system for fast breeder test reactor

    International Nuclear Information System (INIS)

    The objective of this project is to computerize and web-enable the Work Permit System for the Fast Breeder Test Reactor (FBTR) at IGCAR, Kalpakkam. The existing Work Permit System at FBTR was studied in detail. Since all the formalities were paper-based, the risk of human error in scrutinizing all permits before reactor start-up was high. Compilation of reports (daily, monthly, yearly etc.) was tedious. The work permit system was therefore automated in order to enable the operation group manage the maintenance work carried out in the plant systematically with entries. The entire project was classified into five permit modules -maintenance, transfer, return, cancellation and reissue. Each module takes care of the entry and maintenance of data in their respective fields in their respective tables. The user is also provided with an option to take a hard copy of the report of his/her choice. A client/server based system was designed to web-enable the entire project. The server program was designed using VB 6.0 as the front-end and MS Access database as the back end to store the data. The client software was developed using Active Server Pages and published using personal web server in the Intranet. A number of administrative tools have been incorporated in the software to ensure access security and integrity of the database. An online help feature with search facilities was added to the software. The work permit system software is now already being used at FBTR and has been deemed to be an invaluable aid in empowering the availability of the reactor and determining the performance history of the equipment. (author)

  11. Nuclear reactor fuel element

    International Nuclear Information System (INIS)

    The fuel element box for a BWR is situated with a corner bolt on the inside in one corner of its top on the top side of the top plate. This corner bolt is screwed down with a bolt with a corner part which is provided with leaf springs outside on two sides, where the bolt has a smaller diameter and an expansion shank. The bolt is held captive to the bolt head on the top and the holder on the bottom of the corner part. The holder is a locknut. If the expansion forces are too great, the bolt can only break at the expansion shank. (HP)

  12. Knowledge management in fast reactors and related fuel cycles

    International Nuclear Information System (INIS)

    regarding BFS-1, BFS-2 and KOBR and post irradiation experience. In UK a super archive was prepared. In USA, TREAT and ZPPR data are currently on a magnetic tape and hard copies with some transfer to electronic files. It is therefore subject to loss. Hence selected ZPR and ZPPR log books are being scanned and selected critical configurations are being preserved. It is needless to emphasize that in R and D organizations like Indira Gandhi Centre for Atomic Research (IGCAR) with a mandate to conduct broad based multi disciplinary programme of scientific research and advanced engineering directed towards fast reactor technology and associated fuel cycle facilities, knowledge management plays a vital role. It also helps in our vision to achieve world class leadership in the fields of Fast Reactor technology and related Fuel Cycles. Also, India would like to achieve energy security through Fast Breeder Reactors. IGCAR has been operating a Fast Breeder Test Reactor (FBTR) successfully for the last 23 years with a unique Pu-U carbide fuel. The Centre has developed and nurturing world class expertise in the areas of fast reactor engineering, reactor safety and analysis, sodium technology, materials development and characterization, non destructive evaluation, in service inspection, reactor instrumentation, computer modeling etc. The centre had successfully reprocessed the Pu-U carbide fuel from FBTR of 150,000 MWd/t burn up. A lot of knowledge has been created in these domains and is being effectively managed and utilized. Learning from 380 reactor years of knowledge base of international experience and knowledge accrued from our own Fast Breeder Test Reactor through successful operation for 20 years and with major engineering experiments in fast reactor technology conducted, IGCAR has indigenously designed 500 MWe Prototype Fast Breeder Reactor (PFBR) and the reactor is under construction. With creative management of knowledge of the centre a Fast Reactor Fuel Cycle Facility is

  13. Fusion breeder neutronics. Final report

    International Nuclear Information System (INIS)

    Research efforts in fusion breeder neutronics have been focused on two tasks that are strongly related. Efforts in Task 1 concentrate on examining the required conditions to sustain fuel self-sufficiency in fusion reactors operated on a D-T fuel cycle. In this respect, in-depth and detailed engineering analyses have been performed on various blanket and reactor concepts to verify the potential of each blanket concept to exhibit a tritium breeding ratio (TBR) in excess of unity by a margin that compensates for losses, radioactive decay and other inventory requirements. Efforts in Task 2 concentrate on evaluating the overall uncertainties (both experimental and analytical) associated with the TBR

  14. Swiss breeder research programme

    International Nuclear Information System (INIS)

    to find a sound base for the financial support. For the reactor physics research two almost parallel activities were considered. During the first period mainly existing know-how will be applied and a step by step familiarisation with the significance of fast breeder reactor physics is foreseen. New pointwise and group-wise cross section libraries based on ENDFIB-VI and JEF 1.1 have been prepared. A large (1250 MWe) sodium-cooled fast breeder reactor benchmark problem was calculated and the eigenvalues, isothermal core fuel Doppler-reactivities, effective delayed neutron fraction and reactivity worths were compared with a great number of solutions obtained in the past. During the following period new methods and models to calculate burnup-cycles of large breeder reactors should be developed and tested. Data libraries for shielding problems to be used in the ECCO code will be prepared and shielding problems calculated. The thermal hydraulics research is conducted to investigate the flow structures produced by two parallel layers of liquid at different velocities and temperatures. This problem arises particularly on occasions where natural circulation is prevailing and hot and cold streams of liquid come together. At present, tests are carried out with water in an horizontal glass channel (WAMIX). Two flow visualization techniques are being used: laser-sheet induced luminescence and image-analysis of video pictures taken with ink injection marking. Based on the image analysis a determination of the frequency of appearance of vortices (time-dependence) could be made. In the analytical area the computational thermal hydraulics code ASTEC was further validated by participation in an international benchmark calculation exercise. This code is also used to calculate the velocity profiles in the boundary layer of the inlet segment of the WAMIX test section. It is intended to directly participate in the European and the French R and D programmes for sodium-cooled fast breeder

  15. Data management for the Clinch River Breeder Reactor Plant Project by use of document status and hold systems

    International Nuclear Information System (INIS)

    This paper describes the development, framework, and scope of the Document Status System and the Document Hold System for the Clinch River Breeder Reactor Plant Project. It shows how data are generated at five locations and transmitted to a central computer for processing and storage. The resulting computerized data bank provides reports needed to perform day-to-day management and engineering planning. Those reports also partially satisfy the requirements of the Project's Quality Assurance Program

  16. Knowledge Management in Fast Reactors and Related Fuel Cycles

    International Nuclear Information System (INIS)

    The 21st century is ushering in a new phase of economic and social development which can be referred as 'Knowledge Economy' in which knowledge has become the key asset in determining the organization's success or failure. Nuclear technology is very complex and a highly technical endeavor. It relies on innovative creation, storage and dissemination of knowledge. The nuclear technology is also characterized by long time scales and technological excellence. Nuclear Knowledge Management is a critical input to Nuclear Power Industry, the associated fuel cycle activities and nuclear applications in medicine, industry and agriculture. In an R and D Organization like Indira Gandhi Centre for Atomic Research (IGCAR) specializing in the areas of Fast Reactor Technology and associated Fuel Cycle Facilities, Knowledge Management plays a vital role. IGCAR is operating successfully the Fast Breeder Test Reactor (FBTR) for the last 24 years with a unique Pu-U Carbide Fuel. The paper highlights the various success stories, lessons learnt from FBTR, knowledge accrued, disseminated and reused. With intensive R and D and innovations, the processes have been developed and FBTR's spent carbide fuel of 155 GWd/t burn up has been reprocessed successfully. The paper covers the knowledge that has been created through extensive analysis and validation for the design of a 500 MWe Prototype Fast Breeder Reactor (PFBR) which is under construction at Kalpakkam. The Centre has developed world class expertise in the areas of sodium technology, material development, non-destructive evaluation, instrumentation etc. This paper gives some examples of how the knowledge generated is used for PFBR. (author)

  17. Thermophysical properties of reactor fuels

    International Nuclear Information System (INIS)

    A review is presented of the literature on the enthalpy of uranium, thorium, and plutonium oxide and an approach is described for calculating the vapor pressure and gaseous composition of reactor fuel. In these calculations, thermodynamic functions of gas phase molecular species (obtained from matrix-isolation spectroscopy) are employed in conjunction with condensed phase therodynamics. A summary is presented of the status of this work

  18. Stress Analysis of Steam Generator Shell Nozzle Junction for Sodium cooled Fast Breeder Reactor

    Directory of Open Access Journals (Sweden)

    Mani N,

    2010-07-01

    Full Text Available The Steam Generators (SG decides the capacity factor in Sodium cooled Fast breeder Reactor (SFR plants and hence they are designed with high reliability. One of the critical locations in SG is the shell nozzle junction. This junction is subjected to an end bending moment and internal pressure. Since the shell nozzle junction is the critical location of SG a double-ended guillotine rupture will result in leakage of large quantity of sodium, which is not desirable. The material of construction is modified 9Cr-1Mo. Hence safety equirements demand that Leak Before Break criteria with assumed initial flaw have to be demonstrated. To demonstrate LBB, the basic requirement is to predict the state of stress precisely in the shell nozzle junction under various loading conditions. An efficient finiteelement modeling for shell nozzle junction has been presented in which shell elements are employed to idealize the whole region. These results are used for the analysis of leak before break concept.

  19. Techniques developed to evaluate the fracture toughness offast breeder reactor duct

    International Nuclear Information System (INIS)

    Large changes in strength and ductility of metals after irradiation are known to occur. The fracture toughness of irradiated metals, which is related to the combined strength and ductility of a material, may be significantly reduced and the potential for unstable crack extension increased. Therefore, the resistance of cladding and duct materials to fracture after exposure to fast neutron environments is of concern. Existing Type 316 stainless steel irradiated ducts are relatively thin and since this material retains substantial ductility, even after irradiation, the fracture behavior of the duct material cannot be analyzed by linear elastic fracture mechanics techniques. Instead, the multispecimen R-curve method and J-integral analysis were used to develop an experimental approach to evaluate the fracture toughness of thin breeder reactor duct materials irradiated at elevated temperatures. Alloy A-286 was chosen for these experiments because the alloy exhibits elastic/plastic behavior and the fracture toughness data of thicker (12 mm) specimens were available for comparison. Technical problems associated with specimen buckling and remote handling were treated in this work. The results are discussed in terms of thickness criterion for plane strain

  20. Power supply for control and instrumentation in Fast Breeder Test Reactor

    International Nuclear Information System (INIS)

    The design and operation of the four 'no-break' power supplies for control and instrumentation in the Fast Breeder Test Reactor (FBTR), Kalpakkam, are described. Interruptions in the power supplies are eliminated by redundancy and battery back-up source while voltage dips and transients are taken care by automatic regulation system. The four power supplies are : (1) 24 V D.C. exclusively for neutronic and safety circuits, (2) 48 V D.C. for control logic indication lamps and solenoid valves, (3) 220 V D.C. for switchgear control, control room emergency lighting and D.C. flushing oil pump for the turbine and (4) 220 V A.C. single-phase 50 H/Z for computers and electronics of control and instrumentation. Stationary lead-acid batteries (lead antimony type) in floating mode operation with rectifier/charger are used for emergency back-up. All these power supplies are fed by 415 V, 3-phase, 50 HZ emergency supply buses which are provided with diesel generator back-up. Static energy conversion system (in preference to mechanical rotation system) is used for A.C. to D.C. and also for A.C. to A.C. conversion. (M.G.B.)